RELAP-7 Closure Correlations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zou, Ling; Berry, R. A.; Martineau, R. C.

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The code is based on the INL’s modern scientific software development framework, MOOSE (Multi-Physics Object Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5’s and TRACE’s capabilities and extends their analysis capabilities for all reactor system simulation scenarios. The RELAP-7 codemore » utilizes the well-posed 7-equation two-phase flow model for compressible two-phase flow. Closure models used in the TRACE code has been reviewed and selected to reflect the progress made during the past decades and provide a basis for the colure correlations implemented in the RELAP-7 code. This document provides a summary on the closure correlations that are currently implemented in the RELAP-7 code. The closure correlations include sub-grid models that describe interactions between the fluids and the flow channel, and interactions between the two phases.« less

Reactor Application for Coaching Newbies

DOE Office of Scientific and Technical Information (OSTI.GOV)

2015-06-17

RACCOON is a Moose based reactor physics application designed to engage undergraduate and first-year graduate students. The code contains capabilities to solve the multi group Neutron Diffusion equation in eigenvalue and fixed source form and will soon have a provision to provide simple thermal feedback. These capabilities are sufficient to solve example problems found in Duderstadt & Hamilton (the typical textbook of senior level reactor physics classes). RACCOON does not contain any advanced capabilities as found in YAK.

Advances in modelling of condensation phenomena

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liu, W.S.; Zaltsgendler, E.; Hanna, B.

1997-07-01

The physical parameters in the modelling of condensation phenomena in the CANDU reactor system codes are discussed. The experimental programs used for thermal-hydraulic code validation in the Canadian nuclear industry are briefly described. The modelling of vapour generation and in particular condensation plays a key role in modelling of postulated reactor transients. The condensation models adopted in the current state-of-the-art two-fluid CANDU reactor thermal-hydraulic system codes (CATHENA and TUF) are described. As examples of the modelling challenges faced, the simulation of a cold water injection experiment by CATHENA and the simulation of a condensation induced water hammer experiment by TUFmore » are described.« less

Hot zero power reactor calculations using the Insilico code

DOE PAGES

Hamilton, Steven P.; Evans, Thomas M.; Davidson, Gregory G.; ...

2016-03-18

In this paper we describe the reactor physics simulation capabilities of the insilico code. A description of the various capabilities of the code is provided, including detailed discussion of the geometry, meshing, cross section processing, and neutron transport options. Numerical results demonstrate that the insilico SP N solver with pin-homogenized cross section generation is capable of delivering highly accurate full-core simulation of various PWR problems. Comparison to both Monte Carlo calculations and measured plant data is provided.

RELAP-7 Development Updates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, Hongbin; Zhao, Haihua; Gleicher, Frederick Nathan

RELAP-7 is a nuclear systems safety analysis code being developed at the Idaho National Laboratory, and is the next generation tool in the RELAP reactor safety/systems analysis application series. RELAP-7 development began in 2011 to support the Risk Informed Safety Margins Characterization (RISMC) Pathway of the Light Water Reactor Sustainability (LWRS) program. The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical methods, and physical models in order to provide capabilities needed for the RISMC methodology and to support nuclear power safety analysis. The code is beingmore » developed based on Idaho National Laboratory’s modern scientific software development framework – MOOSE (the Multi-Physics Object-Oriented Simulation Environment). The initial development goal of the RELAP-7 approach focused primarily on the development of an implicit algorithm capable of strong (nonlinear) coupling of the dependent hydrodynamic variables contained in the 1-D/2-D flow models with the various 0-D system reactor components that compose various boiling water reactor (BWR) and pressurized water reactor nuclear power plants (NPPs). During Fiscal Year (FY) 2015, the RELAP-7 code has been further improved with expanded capability to support boiling water reactor (BWR) and pressurized water reactor NPPs analysis. The accumulator model has been developed. The code has also been coupled with other MOOSE-based applications such as neutronics code RattleSnake and fuel performance code BISON to perform multiphysics analysis. A major design requirement for the implicit algorithm in RELAP-7 is that it is capable of second-order discretization accuracy in both space and time, which eliminates the traditional first-order approximation errors. The second-order temporal is achieved by a second-order backward temporal difference, and the one-dimensional second-order accurate spatial discretization is achieved with the Galerkin approximation of Lagrange finite elements. During FY-2015, we have done numerical verification work to verify that the RELAP-7 code indeed achieves 2nd-order accuracy in both time and space for single phase models at the system level.« less

The development of a thermal hydraulic feedback mechanism with a quasi-fixed point iteration scheme for control rod position modeling for the TRIGSIMS-TH application

NASA Astrophysics Data System (ADS)

Karriem, Veronica V.

Nuclear reactor design incorporates the study and application of nuclear physics, nuclear thermal hydraulic and nuclear safety. Theoretical models and numerical methods implemented in computer programs are utilized to analyze and design nuclear reactors. The focus of this PhD study's is the development of an advanced high-fidelity multi-physics code system to perform reactor core analysis for design and safety evaluations of research TRIGA-type reactors. The fuel management and design code system TRIGSIMS was further developed to fulfill the function of a reactor design and analysis code system for the Pennsylvania State Breazeale Reactor (PSBR). TRIGSIMS, which is currently in use at the PSBR, is a fuel management tool, which incorporates the depletion code ORIGEN-S (part of SCALE system) and the Monte Carlo neutronics solver MCNP. The diffusion theory code ADMARC-H is used within TRIGSIMS to accelerate the MCNP calculations. It manages the data and fuel isotopic content and stores it for future burnup calculations. The contribution of this work is the development of an improved version of TRIGSIMS, named TRIGSIMS-TH. TRIGSIMS-TH incorporates a thermal hydraulic module based on the advanced sub-channel code COBRA-TF (CTF). CTF provides the temperature feedback needed in the multi-physics calculations as well as the thermal hydraulics modeling capability of the reactor core. The temperature feedback model is using the CTF-provided local moderator and fuel temperatures for the cross-section modeling for ADMARC-H and MCNP calculations. To perform efficient critical control rod calculations, a methodology for applying a control rod position was implemented in TRIGSIMS-TH, making this code system a modeling and design tool for future core loadings. The new TRIGSIMS-TH is a computer program that interlinks various other functional reactor analysis tools. It consists of the MCNP5, ADMARC-H, ORIGEN-S, and CTF. CTF was coupled with both MCNP and ADMARC-H to provide the heterogeneous temperature distribution throughout the core. Each of these codes is written in its own computer language performing its function and outputs a set of data. TRIGSIMS-TH provides an effective use and data manipulation and transfer between different codes. With the implementation of feedback and control- rod-position modeling methodologies, the TRIGSIMS-TH calculations are more accurate and in a better agreement with measured data. The PSBR is unique in many ways and there are no "off-the-shelf" codes, which can model this design in its entirety. In particular, PSBR has an open core design, which is cooled by natural convection. Combining several codes into a unique system brings many challenges. It also requires substantial knowledge of both operation and core design of the PSBR. This reactor is in operation decades and there is a fair amount of studies and developments in both PSBR thermal hydraulics and neutronics. Measured data is also available for various core loadings and can be used for validation activities. The previous studies and developments in PSBR modeling also aids as a guide to assess the findings of the work herein. In order to incorporate new methods and codes into exiting TRIGSIMS, a re-evaluation of various components of the code was performed to assure the accuracy and efficiency of the existing CTF/MCNP5/ADMARC-H multi-physics coupling. A new set of ADMARC-H diffusion coefficients and cross sections was generated using the SERPENT code. This was needed as the previous data was not generated with thermal hydraulic feedback and the ARO position was used as the critical rod position. The B4C was re-evaluated for this update. The data exchange between ADMARC-H and MCNP5 was modified. The basic core model is given a flexibility to allow for various changes within the core model, and this feature was implemented in TRIGSIMS-TH. The PSBR core in the new code model can be expanded and changed. This allows the new code to be used as a modeling tool for design and analyses of future code loadings.

Neutron flux and power in RTP core-15

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rabir, Mohamad Hairie, E-mail: m-hairie@nuclearmalaysia.gov.my; Zin, Muhammad Rawi Md; Usang, Mark Dennis

PUSPATI TRIGA Reactor achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. This paper describes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP); focusing on the application of the developed reactor 3D model for criticality calculation, analysis of power and neutron flux distribution of TRIGA core. The 3D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA reactor. The model represents in detailed all important components of the core withmore » literally no physical approximation. The consistency and accuracy of the developed RTP MCNP model was established by comparing calculations to the available experimental results and TRIGLAV code calculation.« less

SAM Theory Manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Rui

The System Analysis Module (SAM) is an advanced and modern system analysis tool being developed at Argonne National Laboratory under the U.S. DOE Office of Nuclear Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. SAM development aims for advances in physical modeling, numerical methods, and software engineering to enhance its user experience and usability for reactor transient analyses. To facilitate the code development, SAM utilizes an object-oriented application framework (MOOSE), and its underlying meshing and finite-element library (libMesh) and linear and non-linear solvers (PETSc), to leverage modern advanced software environments and numerical methods. SAM focuses on modeling advanced reactormore » concepts such as SFRs (sodium fast reactors), LFRs (lead-cooled fast reactors), and FHRs (fluoride-salt-cooled high temperature reactors) or MSRs (molten salt reactors). These advanced concepts are distinguished from light-water reactors in their use of single-phase, low-pressure, high-temperature, and low Prandtl number (sodium and lead) coolants. As a new code development, the initial effort has been focused on modeling and simulation capabilities of heat transfer and single-phase fluid dynamics responses in Sodium-cooled Fast Reactor (SFR) systems. The system-level simulation capabilities of fluid flow and heat transfer in general engineering systems and typical SFRs have been verified and validated. This document provides the theoretical and technical basis of the code to help users understand the underlying physical models (such as governing equations, closure models, and component models), system modeling approaches, numerical discretization and solution methods, and the overall capabilities in SAM. As the code is still under ongoing development, this SAM Theory Manual will be updated periodically to keep it consistent with the state of the development.« less

Analysis of JSI TRIGA MARK II reactor physical parameters calculated with TRIPOLI and MCNP.

PubMed

Henry, R; Tiselj, I; Snoj, L

2015-03-01

New computational model of the JSI TRIGA Mark II research reactor was built for TRIPOLI computer code and compared with existing MCNP code model. The same modelling assumptions were used in order to check the differences of the mathematical models of both Monte Carlo codes. Differences between the TRIPOLI and MCNP predictions of keff were up to 100pcm. Further validation was performed with analyses of the normalized reaction rates and computations of kinetic parameters for various core configurations. Copyright © 2014 Elsevier Ltd. All rights reserved.

Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

NASA Astrophysics Data System (ADS)

Antariksawan, Anhar R.; Wahyono, Puradwi I.; Taxwim

2018-02-01

Safety is the priority for nuclear installations, including research reactors. On the other hand, many studies have been done to validate the applicability of nuclear power plant based best estimate computer codes to the research reactor. This study aims to assess the applicability of the RELAP5/SCDAP code to Kartini research reactor. The model development, steady state and transient due to LOCA calculations have been conducted by using RELAP5/SCDAP. The calculation results are compared with available measurements data from Kartini research reactor. The results show that the RELAP5/SCDAP model steady state calculation agrees quite well with the available measurement data. While, in the case of LOCA transient simulations, the model could result in reasonable physical phenomena during the transient showing the characteristics and performances of the reactor against the LOCA transient. The role of siphon breaker hole and natural circulation in the reactor tank as passive system was important to keep reactor in safe condition. It concludes that the RELAP/SCDAP could be use as one of the tool to analyse the thermal-hydraulic safety of Kartini reactor. However, further assessment to improve the model is still needed.

Nuclear thermal propulsion engine system design analysis code development

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.; Ivanenok, Joseph F.

1992-01-01

A Nuclear Thermal Propulsion (NTP) Engine System Design Analyis Code has recently been developed to characterize key NTP engine system design features. Such a versatile, standalone NTP system performance and engine design code is required to support ongoing and future engine system and vehicle design efforts associated with proposed Space Exploration Initiative (SEI) missions of interest. Key areas of interest in the engine system modeling effort were the reactor, shielding, and inclusion of an engine multi-redundant propellant pump feed system design option. A solid-core nuclear thermal reactor and internal shielding code model was developed to estimate the reactor's thermal-hydraulic and physical parameters based on a prescribed thermal output which was integrated into a state-of-the-art engine system design model. The reactor code module has the capability to model graphite, composite, or carbide fuels. Key output from the model consists of reactor parameters such as thermal power, pressure drop, thermal profile, and heat generation in cooled structures (reflector, shield, and core supports), as well as the engine system parameters such as weight, dimensions, pressures, temperatures, mass flows, and performance. The model's overall analysis methodology and its key assumptions and capabilities are summarized in this paper.

Towards a supported common NEAMS software stack

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cormac Garvey

2012-04-01

The NEAMS IPSC's are developing multidimensional, multiphysics, multiscale simulation codes based on first principles that will be capable of predicting all aspects of current and future nuclear reactor systems. These new breeds of simulation codes will include rigorous verification, validation and uncertainty quantification checks to quantify the accuracy and quality of the simulation results. The resulting NEAMS IPSC simulation codes will be an invaluable tool in designing the next generation of Nuclear Reactors and also contribute to a more speedy process in the acquisition of licenses from the NRC for new Reactor designs. Due to the high resolution of themore » models, the complexity of the physics and the added computational resources to quantify the accuracy/quality of the results, the NEAMS IPSC codes will require large HPC resources to carry out the production simulation runs.« less

a Dosimetry Assessment for the Core Restraint of AN Advanced Gas Cooled Reactor

NASA Astrophysics Data System (ADS)

Thornton, D. A.; Allen, D. A.; Tyrrell, R. J.; Meese, T. C.; Huggon, A. P.; Whiley, G. S.; Mossop, J. R.

2009-08-01

This paper describes calculations of neutron damage rates within the core restraint structures of Advanced Gas Cooled Reactors (AGRs). Using advanced features of the Monte Carlo radiation transport code MCBEND, and neutron source data from core follow calculations performed with the reactor physics code PANTHER, a detailed model of the reactor cores of two of British Energy's AGR power plants has been developed for this purpose. Because there are no relevant neutron fluence measurements directly supporting this assessment, results of benchmark comparisons and successful validation of MCBEND for Magnox reactors have been used to estimate systematic and random uncertainties on the predictions. In particular, it has been necessary to address the known under-prediction of lower energy fast neutron responses associated with the penetration of large thicknesses of graphite.

High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems.

PubMed

Mahadevan, Vijay S; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul

2014-08-06

An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework.

High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems

PubMed Central

Mahadevan, Vijay S.; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul

2014-01-01

An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework. PMID:24982250

DOE Office of Scientific and Technical Information (OSTI.GOV)

Davis, R. S.

The following are specific topics of this paper: 1.There is much creativity in the manner in which Dimensional Generator can be applied to a specific programming task [2]. This paper tells how Dimensional Generator was applied to a reactor-physics task. 2. In this first practical use, Dimensional Generator itself proved not to need change, but a better user interface was found necessary, essentially because the relevance of Dimensional Generator to reactor physics was initially underestimated. It is briefly described. 3. The use of Dimensional Generator helps make reactor-physics source code somewhat simpler. That is explained here with brief examples frommore » BURFEL-PC and WIMSBURF. 4. Most importantly, with the help of Dimensional Generator, all erroneous physical expressions were automatically detected. The errors are detailed here (in spite of the author's embarrassment) because they show clearly, both in theory and in practice, how Dimensional Generator offers quality enhancement of reactor-physics programming. (authors)« less

Reactivity effects in VVER-1000 of the third unit of the kalinin nuclear power plant at physical start-up. Computations in ShIPR intellectual code system with library of two-group cross sections generated by UNK code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zizin, M. N.; Zimin, V. G.; Zizina, S. N., E-mail: zizin@adis.vver.kiae.ru

2010-12-15

The ShIPR intellectual code system for mathematical simulation of nuclear reactors includes a set of computing modules implementing the preparation of macro cross sections on the basis of the two-group library of neutron-physics cross sections obtained for the SKETCH-N nodal code. This library is created by using the UNK code for 3D diffusion computation of first VVER-1000 fuel loadings. Computation of neutron fields in the ShIPR system is performed using the DP3 code in the two-group diffusion approximation in 3D triangular geometry. The efficiency of all groups of control rods for the first fuel loading of the third unit ofmore » the Kalinin Nuclear Power Plant is computed. The temperature, barometric, and density effects of reactivity as well as the reactivity coefficient due to the concentration of boric acid in the reactor were computed additionally. Results of computations are compared with the experiment.« less

Reactivity effects in VVER-1000 of the third unit of the kalinin nuclear power plant at physical start-up. Computations in ShIPR intellectual code system with library of two-group cross sections generated by UNK code

NASA Astrophysics Data System (ADS)

Zizin, M. N.; Zimin, V. G.; Zizina, S. N.; Kryakvin, L. V.; Pitilimov, V. A.; Tereshonok, V. A.

2010-12-01

The ShIPR intellectual code system for mathematical simulation of nuclear reactors includes a set of computing modules implementing the preparation of macro cross sections on the basis of the two-group library of neutron-physics cross sections obtained for the SKETCH-N nodal code. This library is created by using the UNK code for 3D diffusion computation of first VVER-1000 fuel loadings. Computation of neutron fields in the ShIPR system is performed using the DP3 code in the two-group diffusion approximation in 3D triangular geometry. The efficiency of all groups of control rods for the first fuel loading of the third unit of the Kalinin Nuclear Power Plant is computed. The temperature, barometric, and density effects of reactivity as well as the reactivity coefficient due to the concentration of boric acid in the reactor were computed additionally. Results of computations are compared with the experiment.

Vectorized Monte Carlo methods for reactor lattice analysis

NASA Technical Reports Server (NTRS)

Brown, F. B.

1984-01-01

Some of the new computational methods and equivalent mathematical representations of physics models used in the MCV code, a vectorized continuous-enery Monte Carlo code for use on the CYBER-205 computer are discussed. While the principal application of MCV is the neutronics analysis of repeating reactor lattices, the new methods used in MCV should be generally useful for vectorizing Monte Carlo for other applications. For background, a brief overview of the vector processing features of the CYBER-205 is included, followed by a discussion of the fundamentals of Monte Carlo vectorization. The physics models used in the MCV vectorized Monte Carlo code are then summarized. The new methods used in scattering analysis are presented along with details of several key, highly specialized computational routines. Finally, speedups relative to CDC-7600 scalar Monte Carlo are discussed.

Validation of the WIMSD4M cross-section generation code with benchmark results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Deen, J.R.; Woodruff, W.L.; Leal, L.E.

1995-01-01

The WIMSD4 code has been adopted for cross-section generation in support of the Reduced Enrichment Research and Test Reactor (RERTR) program at Argonne National Laboratory (ANL). Subsequently, the code has undergone several updates, and significant improvements have been achieved. The capability of generating group-collapsed micro- or macroscopic cross sections from the ENDF/B-V library and the more recent evaluation, ENDF/B-VI, in the ISOTXS format makes the modified version of the WIMSD4 code, WIMSD4M, very attractive, not only for the RERTR program, but also for the reactor physics community. The intent of the present paper is to validate the WIMSD4M cross-section librariesmore » for reactor modeling of fresh water moderated cores. The results of calculations performed with multigroup cross-section data generated with the WIMSD4M code will be compared against experimental results. These results correspond to calculations carried out with thermal reactor benchmarks of the Oak Ridge National Laboratory (ORNL) unreflected HEU critical spheres, the TRX LEU critical experiments, and calculations of a modified Los Alamos HEU D{sub 2}O moderated benchmark critical system. The benchmark calculations were performed with the discrete-ordinates transport code, TWODANT, using WIMSD4M cross-section data. Transport calculations using the XSDRNPM module of the SCALE code system are also included. In addition to transport calculations, diffusion calculations with the DIF3D code were also carried out, since the DIF3D code is used in the RERTR program for reactor analysis and design. For completeness, Monte Carlo results of calculations performed with the VIM and MCNP codes are also presented.« less

Optimization and parallelization of the thermal–hydraulic subchannel code CTF for high-fidelity multi-physics applications

DOE PAGES

Salko, Robert K.; Schmidt, Rodney C.; Avramova, Maria N.

2014-11-23

This study describes major improvements to the computational infrastructure of the CTF subchannel code so that full-core, pincell-resolved (i.e., one computational subchannel per real bundle flow channel) simulations can now be performed in much shorter run-times, either in stand-alone mode or as part of coupled-code multi-physics calculations. These improvements support the goals of the Department Of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL) Energy Innovation Hub to develop high fidelity multi-physics simulation tools for nuclear energy design and analysis.

Calculated criticality for sup 235 U/graphite systems using the VIM Monte Carlo code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collins, P.J.; Grasseschi, G.L.; Olsen, D.N.

1992-01-01

Calculations for highly enriched uranium and graphite systems gained renewed interest recently for the new production modular high-temperature gas-cooled reactor (MHTGR). Experiments to validate the physics calculations for these systems are being prepared for the Transient Reactor Test Facility (TREAT) reactor at Argonne National Laboratory (ANL-West) and in the Compact Nuclear Power Source facility at Los Alamos National Laboratory. The continuous-energy Monte Carlo code VIM, or equivalently the MCNP code, can utilize fully detailed models of the MHTGR and serve as benchmarks for the approximate multigroup methods necessary in full reactor calculations. Validation of these codes and their associated nuclearmore » data did not exist for highly enriched {sup 235}U/graphite systems. Experimental data, used in development of more approximate methods, dates back to the 1960s. The authors have selected two independent sets of experiments for calculation with the VIM code. The carbon-to-uranium (C/U) ratios encompass the range of 2,000, representative of the new production MHTGR, to the ratio of 10,000 in the fuel of TREAT. Calculations used the ENDF/B-V data.« less

Monte Carol-based validation of neutronic methodology for EBR-II analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liaw, J.R.; Finck, P.J.

1993-01-01

The continuous-energy Monte Carlo code VIM (Ref. 1) has been validated extensively over the years against fast critical experiments and other neutronic analysis codes. A high degree of confidence in VIM for predicting reactor physics parameters has been firmly established. This paper presents a numerical validation of two conventional multigroup neutronic analysis codes, DIF3D (Ref. 4) and VARIANT (Ref. 5), against VIM for two Experimental Breeder Reactor II (EBR-II) core loadings in detailed three-dimensional hexagonal-z geometry. The DIF3D code is based on nodal diffusion theory, and it is used in calculations for day-today reactor operations, whereas the VARIANT code ismore » based on nodal transport theory and is used with increasing frequency for specific applications. Both DIF3D and VARIANT rely on multigroup cross sections generated from ENDF/B-V by the ETOE-2/MC[sup 2]-II/SDX (Ref. 6) code package. Hence, this study also validates the multigroup cross-section processing methodology against the continuous-energy approach used in VIM.« less

High-Fidelity Coupled Monte-Carlo/Thermal-Hydraulics Calculations

NASA Astrophysics Data System (ADS)

Ivanov, Aleksandar; Sanchez, Victor; Ivanov, Kostadin

2014-06-01

Monte Carlo methods have been used as reference reactor physics calculation tools worldwide. The advance in computer technology allows the calculation of detailed flux distributions in both space and energy. In most of the cases however, those calculations are done under the assumption of homogeneous material density and temperature distributions. The aim of this work is to develop a consistent methodology for providing realistic three-dimensional thermal-hydraulic distributions by coupling the in-house developed sub-channel code SUBCHANFLOW with the standard Monte-Carlo transport code MCNP. In addition to the innovative technique of on-the fly material definition, a flux-based weight-window technique has been introduced to improve both the magnitude and the distribution of the relative errors. Finally, a coupled code system for the simulation of steady-state reactor physics problems has been developed. Besides the problem of effective feedback data interchange between the codes, the treatment of temperature dependence of the continuous energy nuclear data has been investigated.

FERRET data analysis code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schmittroth, F.

1979-09-01

A documentation of the FERRET data analysis code is given. The code provides a way to combine related measurements and calculations in a consistent evaluation. Basically a very general least-squares code, it is oriented towards problems frequently encountered in nuclear data and reactor physics. A strong emphasis is on the proper treatment of uncertainties and correlations and in providing quantitative uncertainty estimates. Documentation includes a review of the method, structure of the code, input formats, and examples.

REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nichols, T.; Beals, D.; Sternat, M.

2011-07-18

Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Manymore » research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical analysis. The fuel assembly was modeled using MONTEBURNS(MCNP5/ ORIGEN2.2) and MCNPX/CINDER90. The results from the models have been compared to each other and to the measured data.« less

Hybrid parallel code acceleration methods in full-core reactor physics calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Courau, T.; Plagne, L.; Ponicot, A.

2012-07-01

When dealing with nuclear reactor calculation schemes, the need for three dimensional (3D) transport-based reference solutions is essential for both validation and optimization purposes. Considering a benchmark problem, this work investigates the potential of discrete ordinates (Sn) transport methods applied to 3D pressurized water reactor (PWR) full-core calculations. First, the benchmark problem is described. It involves a pin-by-pin description of a 3D PWR first core, and uses a 8-group cross-section library prepared with the DRAGON cell code. Then, a convergence analysis is performed using the PENTRAN parallel Sn Cartesian code. It discusses the spatial refinement and the associated angular quadraturemore » required to properly describe the problem physics. It also shows that initializing the Sn solution with the EDF SPN solver COCAGNE reduces the number of iterations required to converge by nearly a factor of 6. Using a best estimate model, PENTRAN results are then compared to multigroup Monte Carlo results obtained with the MCNP5 code. Good consistency is observed between the two methods (Sn and Monte Carlo), with discrepancies that are less than 25 pcm for the k{sub eff}, and less than 2.1% and 1.6% for the flux at the pin-cell level and for the pin-power distribution, respectively. (authors)« less

High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems

DOE PAGES

Mahadevan, Vijay S.; Merzari, Elia; Tautges, Timothy; ...

2014-06-30

An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in ordermore » to reduce the overall numerical uncertainty while leveraging available computational resources. Finally, the coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework.« less

Modelling of the anti-neutrino production and spectra from a Magnox reactor

NASA Astrophysics Data System (ADS)

Mills, Robert W.; Mountford, David J.; Coleman, Jonathon P.; Metelko, Carl; Murdoch, Matthew; Schnellbach, Yan-Jie

2018-01-01

The anti-neutrino source properties of a fission reactor are governed by the production and beta decay of the radionuclides present and the summation of their individual anti-neutrino spectra. The fission product radionuclide production changes during reactor operation and different fissioning species give rise to different product distributions. It is thus possible to determine some details of reactor operation, such as power, from the anti-neutrino emission to confirm safeguards records. Also according to some published calculations, it may be feasible to observe different anti-neutrino spectra depending on the fissile contents of the reactor fuel and thus determine the reactor's fissile material inventory during operation which could considerable improve safeguards. In mid-2014 the University of Liverpool deployed a prototype anti-neutrino detector at the Wylfa R1 station in Anglesey, United Kingdom based upon plastic scintillator technology developed for the T2K project. The deployment was used to develop the detector electronics and software until the reactor was finally shutdown in December 2015. To support the development of this detector technology for reactor monitoring and to understand its capabilities, the National Nuclear Laboratory modelled this graphite moderated and natural uranium fuelled reactor with existing codes used to support Magnox reactor operations and waste management. The 3D multi-physics code PANTHER was used to determine the individual powers of each fuel element (8×6152) during the year and a half period of monitoring based upon reactor records. The WIMS/TRAIL/FISPIN code route was then used to determine the radionuclide inventory of each nuclide on a daily basis in each element. These nuclide inventories were then used with the BTSPEC code to determine the anti-neutrino spectra and source strength using JEFF-3.1.1 data. Finally the anti-neutrino source from the reactor for each day during the year and a half of monitored reactor operation was calculated. The results of the preliminary calculations are shown and limitations in the methods and data discussed.

RADSOURCE. Volume 1, Part 1, A scaling factor prediction computer program technical manual and code validation: Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vance, J.N.; Holderness, J.H.; James, D.W.

1992-12-01

Waste stream scaling factors based on sampling programs are vulnerable to one or more of the following factors: sample representativeness, analytic accuracy, and measurement sensitivity. As an alternative to sample analyses or as a verification of the sampling results, this project proposes the use of the RADSOURCE code, which accounts for the release of fuel-source radionuclides. Once the release rates of these nuclides from fuel are known, the code develops scaling factors for waste streams based on easily measured Cobalt-60 (Co-60) and Cesium-137 (Cs-137). The project team developed mathematical models to account for the appearance rate of 10CFR61 radionuclides inmore » reactor coolant. They based these models on the chemistry and nuclear physics of the radionuclides involved. Next, they incorporated the models into a computer code that calculates plant waste stream scaling factors based on reactor coolant gamma- isotopic data. Finally, the team performed special sampling at 17 reactors to validate the models in the RADSOURCE code.« less

Core follow calculation with the nTRACER numerical reactor and verification using power reactor measurement data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jung, Y. S.; Joo, H. G.; Yoon, J. I.

The nTRACER direct whole core transport code employing the planar MOC solution based 3-D calculation method, the subgroup method for resonance treatment, the Krylov matrix exponential method for depletion, and a subchannel thermal/hydraulic calculation solver was developed for practical high-fidelity simulation of power reactors. Its accuracy and performance is verified by comparing with the measurement data obtained for three pressurized water reactor cores. It is demonstrated that accurate and detailed multi-physic simulation of power reactors is practically realizable without any prior calculations or adjustments. (authors)

Validation of the WIMSD4M cross-section generation code with benchmark results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leal, L.C.; Deen, J.R.; Woodruff, W.L.

1995-02-01

The WIMSD4 code has been adopted for cross-section generation in support of the Reduced Enrichment for Research and Test (RERTR) program at Argonne National Laboratory (ANL). Subsequently, the code has undergone several updates, and significant improvements have been achieved. The capability of generating group-collapsed micro- or macroscopic cross sections from the ENDF/B-V library and the more recent evaluation, ENDF/B-VI, in the ISOTXS format makes the modified version of the WIMSD4 code, WIMSD4M, very attractive, not only for the RERTR program, but also for the reactor physics community. The intent of the present paper is to validate the procedure to generatemore » cross-section libraries for reactor analyses and calculations utilizing the WIMSD4M code. To do so, the results of calculations performed with group cross-section data generated with the WIMSD4M code will be compared against experimental results. These results correspond to calculations carried out with thermal reactor benchmarks of the Oak Ridge National Laboratory(ORNL) unreflected critical spheres, the TRX critical experiments, and calculations of a modified Los Alamos highly-enriched heavy-water moderated benchmark critical system. The benchmark calculations were performed with the discrete-ordinates transport code, TWODANT, using WIMSD4M cross-section data. Transport calculations using the XSDRNPM module of the SCALE code system are also included. In addition to transport calculations, diffusion calculations with the DIF3D code were also carried out, since the DIF3D code is used in the RERTR program for reactor analysis and design. For completeness, Monte Carlo results of calculations performed with the VIM and MCNP codes are also presented.« less

RELAP-7 Level 2 Milestone Report: Demonstration of a Steady State Single Phase PWR Simulation with RELAP-7

DOE Office of Scientific and Technical Information (OSTI.GOV)

David Andrs; Ray Berry; Derek Gaston

The document contains the simulation results of a steady state model PWR problem with the RELAP-7 code. The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on INL's modern scientific software development framework - MOOSE (Multi-Physics Object-Oriented Simulation Environment). This report summarizes the initial results of simulating a model steady-state single phase PWR problem using the current version of the RELAP-7 code. The major purpose of this demonstration simulation is to show that RELAP-7 code can be rapidly developed to simulate single-phase reactor problems. RELAP-7more » is a new project started on October 1st, 2011. It will become the main reactor systems simulation toolkit for RISMC (Risk Informed Safety Margin Characterization) and the next generation tool in the RELAP reactor safety/systems analysis application series (the replacement for RELAP5). The key to the success of RELAP-7 is the simultaneous advancement of physical models, numerical methods, and software design while maintaining a solid user perspective. Physical models include both PDEs (Partial Differential Equations) and ODEs (Ordinary Differential Equations) and experimental based closure models. RELAP-7 will eventually utilize well posed governing equations for multiphase flow, which can be strictly verified. Closure models used in RELAP5 and newly developed models will be reviewed and selected to reflect the progress made during the past three decades. RELAP-7 uses modern numerical methods, which allow implicit time integration, higher order schemes in both time and space, and strongly coupled multi-physics simulations. RELAP-7 is written with object oriented programming language C++. Its development follows modern software design paradigms. The code is easy to read, develop, maintain, and couple with other codes. Most importantly, the modern software design allows the RELAP-7 code to evolve with time. RELAP-7 is a MOOSE-based application. MOOSE (Multiphysics Object-Oriented Simulation Environment) is a framework for solving computational engineering problems in a well-planned, managed, and coordinated way. By leveraging millions of lines of open source software packages, such as PETSC (a nonlinear solver developed at Argonne National Laboratory) and LibMesh (a Finite Element Analysis package developed at University of Texas), MOOSE significantly reduces the expense and time required to develop new applications. Numerical integration methods and mesh management for parallel computation are provided by MOOSE. Therefore RELAP-7 code developers only need to focus on physics and user experiences. By using the MOOSE development environment, RELAP-7 code is developed by following the same modern software design paradigms used for other MOOSE development efforts. There are currently over 20 different MOOSE based applications ranging from 3-D transient neutron transport, detailed 3-D transient fuel performance analysis, to long-term material aging. Multi-physics and multiple dimensional analyses capabilities can be obtained by coupling RELAP-7 and other MOOSE based applications and by leveraging with capabilities developed by other DOE programs. This allows restricting the focus of RELAP-7 to systems analysis-type simulations and gives priority to retain and significantly extend RELAP5's capabilities.« less

Mechanistic prediction of fission-gas behavior during in-cell transient heating tests on LWR fuel using the GRASS-SST and FASTGRASS computer codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rest, J; Gehl, S M

1979-01-01

GRASS-SST and FASTGRASS are mechanistic computer codes for predicting fission-gas behavior in UO/sub 2/-base fuels during steady-state and transient conditions. FASTGRASS was developed in order to satisfy the need for a fast-running alternative to GRASS-SST. Althrough based on GRASS-SST, FASTGRASS is approximately an order of magnitude quicker in execution. The GRASS-SST transient analysis has evolved through comparisons of code predictions with the fission-gas release and physical phenomena that occur during reactor operation and transient direct-electrical-heating (DEH) testing of irradiated light-water reactor fuel. The FASTGRASS calculational procedure is described in this paper, along with models of key physical processes included inmore » both FASTGRASS and GRASS-SST. Predictions of fission-gas release obtained from GRASS-SST and FASTGRASS analyses are compared with experimental observations from a series of DEH tests. The major conclusions is that the computer codes should include an improved model for the evolution of the grain-edge porosity.« less

Development and preliminary verification of the 3D core neutronic code: COCO

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lu, H.; Mo, K.; Li, W.

As the recent blooming economic growth and following environmental concerns (China)) is proactively pushing forward nuclear power development and encouraging the tapping of clean energy. Under this situation, CGNPC, as one of the largest energy enterprises in China, is planning to develop its own nuclear related technology in order to support more and more nuclear plants either under construction or being operation. This paper introduces the recent progress in software development for CGNPC. The focus is placed on the physical models and preliminary verification results during the recent development of the 3D Core Neutronic Code: COCO. In the COCO code,more » the non-linear Green's function method is employed to calculate the neutron flux. In order to use the discontinuity factor, the Neumann (second kind) boundary condition is utilized in the Green's function nodal method. Additionally, the COCO code also includes the necessary physical models, e.g. single-channel thermal-hydraulic module, burnup module, pin power reconstruction module and cross-section interpolation module. The preliminary verification result shows that the COCO code is sufficient for reactor core design and analysis for pressurized water reactor (PWR). (authors)« less

The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 1: Basic Models

NASA Astrophysics Data System (ADS)

Mosunova, N. A.

2018-05-01

The article describes the basic models included in the EUCLID/V1 integrated code intended for safety analysis of liquid metal (sodium, lead, and lead-bismuth) cooled fast reactors using fuel rods with a gas gap and pellet dioxide, mixed oxide or nitride uranium-plutonium fuel under normal operation, under anticipated operational occurrences and accident conditions by carrying out interconnected thermal-hydraulic, neutronics, and thermal-mechanical calculations. Information about the Russian and foreign analogs of the EUCLID/V1 integrated code is given. Modeled objects, equation systems in differential form solved in each module of the EUCLID/V1 integrated code (the thermal-hydraulic, neutronics, fuel rod analysis module, and the burnup and decay heat calculation modules), the main calculated quantities, and also the limitations on application of the code are presented. The article also gives data on the scope of functions performed by the integrated code's thermal-hydraulic module, using which it is possible to describe both one- and twophase processes occurring in the coolant. It is shown that, owing to the availability of the fuel rod analysis module in the integrated code, it becomes possible to estimate the performance of fuel rods in different regimes of the reactor operation. It is also shown that the models implemented in the code for calculating neutron-physical processes make it possible to take into account the neutron field distribution over the fuel assembly cross section as well as other features important for the safety assessment of fast reactors.

VICTORIA: A mechanistic model for radionuclide behavior in the reactor coolant system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schaperow, J.H.; Bixler, N.E.

1996-12-31

VICTORIA is the U.S. Nuclear Regulatory Commission`s (NRC`s) mechanistic, best-estimate code for analysis of fission product release from the core and subsequent transport in the reactor vessel and reactor coolant system. VICTORIA requires thermal-hydraulic data (i.e., temperatures, pressures, and velocities) as input. In the past, these data have been taken from the results of calculations from thermal-hydraulic codes such as SCDAP/RELAP5, MELCOR, and MAAP. Validation and assessment of VICTORIA 1.0 have been completed. An independent peer review of VICTORIA, directed by Brookhaven National Laboratory and supported by experts in the areas of fuel release, fission product chemistry, and aerosol physics,more » has been undertaken. This peer review, which will independently assess the code`s capabilities, is nearing completion with the peer review committee`s final report expected in Dec 1996. A limited amount of additional development is expected as a result of the peer review. Following this additional development, the NRC plans to release VICTORIA 1.1 and an updated and improved code manual. Future plans mainly involve use of the code for plant calculations to investigate specific safety issues as they arise. Also, the code will continue to be used in support of the Phebus experiments.« less

Benchmark Simulation of Natural Circulation Cooling System with Salt Working Fluid Using SAM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahmed, K. K.; Scarlat, R. O.; Hu, R.

Liquid salt-cooled reactors, such as the Fluoride Salt-Cooled High-Temperature Reactor (FHR), offer passive decay heat removal through natural circulation using Direct Reactor Auxiliary Cooling System (DRACS) loops. The behavior of such systems should be well-understood through performance analysis. The advanced system thermal-hydraulics tool System Analysis Module (SAM) from Argonne National Laboratory has been selected for this purpose. The work presented here is part of a larger study in which SAM modeling capabilities are being enhanced for the system analyses of FHR or Molten Salt Reactors (MSR). Liquid salt thermophysical properties have been implemented in SAM, as well as properties ofmore » Dowtherm A, which is used as a simulant fluid for scaled experiments, for future code validation studies. Additional physics modules to represent phenomena specific to salt-cooled reactors, such as freezing of coolant, are being implemented in SAM. This study presents a useful first benchmark for the applicability of SAM to liquid salt-cooled reactors: it provides steady-state and transient comparisons for a salt reactor system. A RELAP5-3D model of the Mark-1 Pebble-Bed FHR (Mk1 PB-FHR), and in particular its DRACS loop for emergency heat removal, provides steady state and transient results for flow rates and temperatures in the system that are used here for code-to-code comparison with SAM. The transient studied is a loss of forced circulation with SCRAM event. To the knowledge of the authors, this is the first application of SAM to FHR or any other molten salt reactors. While building these models in SAM, any gaps in the code’s capability to simulate such systems are identified and addressed immediately, or listed as future improvements to the code.« less

Three-dimensional fuel pin model validation by prediction of hydrogen distribution in cladding and comparison with experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aly, A.; Avramova, Maria; Ivanov, Kostadin

To correctly describe and predict this hydrogen distribution there is a need for multi-physics coupling to provide accurate three-dimensional azimuthal, radial, and axial temperature distributions in the cladding. Coupled high-fidelity reactor-physics codes with a sub-channel code as well as with a computational fluid dynamics (CFD) tool have been used to calculate detailed temperature distributions. These high-fidelity coupled neutronics/thermal-hydraulics code systems are coupled further with the fuel-performance BISON code with a kernel (module) for hydrogen. Both hydrogen migration and precipitation/dissolution are included in the model. Results from this multi-physics analysis is validated utilizing calculations of hydrogen distribution using models informed bymore » data from hydrogen experiments and PIE data.« less

Coupled Physics Environment (CouPE) library - Design, Implementation, and Release

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mahadevan, Vijay S.

Over several years, high fidelity, validated mono-­physics solvers with proven scalability on peta-­scale architectures have been developed independently. Based on a unified component-­based architecture, these existing codes can be coupled with a unified mesh-­data backplane and a flexible coupling-­strategy-­based driver suite to produce a viable tool for analysts. In this report, we present details on the design decisions and developments on CouPE, an acronym that stands for Coupled Physics Environment that orchestrates a coupled physics solver through the interfaces exposed by MOAB array-­based unstructured mesh, both of which are part of SIGMA (Scalable Interfaces for Geometry and Mesh-­Based Applications) toolkit.more » The SIGMA toolkit contains libraries that enable scalable geometry and unstructured mesh creation and handling in a memory and computationally efficient implementation. The CouPE version being prepared for a full open-­source release along with updated documentation will contain several useful examples that will enable users to start developing their applications natively using the native MOAB mesh and couple their models to existing physics applications to analyze and solve real world problems of interest. An integrated multi-­physics simulation capability for the design and analysis of current and future nuclear reactor models is also being investigated as part of the NEAMS RPL, to tightly couple neutron transport, thermal-­hydraulics and structural mechanics physics under the SHARP framework. This report summarizes the efforts that have been invested in CouPE to bring together several existing physics applications namely PROTEUS (neutron transport code), Nek5000 (computational fluid-dynamics code) and Diablo (structural mechanics code). The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The design of CouPE along with motivations that led to implementation choices are also discussed. The first release of the library will be different from the current version of the code that integrates the components in SHARP and explanation on the need for forking the source base will also be provided. Enhancements in the functionality and improved user guides will be available as part of the release. CouPE v0.1 is scheduled for an open-­source release in December 2014 along with SIGMA v1.1 components that provide support for language-agnostic mesh loading, traversal and query interfaces along with scalable solution transfer of fields between different physics codes. The coupling methodology and software interfaces of the library are presented, along with verification studies on two representative fast sodium-­cooled reactor demonstration problems to prove the usability of the CouPE library.« less

Interface requirements to couple thermal-hydraulic codes to severe accident codes: ATHLET-CD

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trambauer, K.

1997-07-01

The system code ATHLET-CD is being developed by GRS in cooperation with IKE and IPSN. Its field of application comprises the whole spectrum of leaks and large breaks, as well as operational and abnormal transients for LWRs and VVERs. At present the analyses cover the in-vessel thermal-hydraulics, the early phases of core degradation, as well as fission products and aerosol release from the core and their transport in the Reactor Coolant System. The aim of the code development is to extend the simulation of core degradation up to failure of the reactor pressure vessel and to cover all physically reasonablemore » accident sequences for western and eastern LWRs including RMBKs. The ATHLET-CD structure is highly modular in order to include a manifold spectrum of models and to offer an optimum basis for further development. The code consists of four general modules to describe the reactor coolant system thermal-hydraulics, the core degradation, the fission product core release, and fission product and aerosol transport. Each general module consists of some basic modules which correspond to the process to be simulated or to its specific purpose. Besides the code structure based on the physical modelling, the code follows four strictly separated steps during the course of a calculation: (1) input of structure, geometrical data, initial and boundary condition, (2) initialization of derived quantities, (3) steady state calculation or input of restart data, and (4) transient calculation. In this paper, the transient solution method is briefly presented and the coupling methods are discussed. Three aspects have to be considered for the coupling of different modules in one code system. First is the conservation of masses and energy in the different subsystems as there are fluid, structures, and fission products and aerosols. Second is the convergence of the numerical solution and stability of the calculation. The third aspect is related to the code performance, and running time.« less

Control rod calibration and reactivity effects at the IPEN/MB-01 reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pinto, Letícia Negrão; Gonnelli, Eduardo; Santos, Adimir dos

2014-11-11

Researches that aim to improve the performance of neutron transport codes and quality of nuclear cross section databases are very important to increase the accuracy of simulations and the quality of the analysis and prediction of phenomena in the nuclear field. In this context, relevant experimental data such as reactivity worth measurements are needed. Control rods may be made of several neutron absorbing materials that are used to adjust the reactivity of the core. For the reactor operation, these experimental data are also extremely important: with them it is possible to estimate the reactivity worth by the movement of themore » control rod, understand the reactor response at each rod position and to operate the reactor safely. This work presents a temperature correction approach for the control rod calibration problem. It is shown the control rod calibration data of the IPEN/MB-01 reactor, the integral and differential reactivity curves and a theoretical analysis, performed by the MCNP-5 reactor physics code, developed and maintained by Los Alamos National Laboratory, using the ENDF/B-VII.0 nuclear data library.« less

Development of 3D pseudo pin-by-pin calculation methodology in ANC

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhang, B.; Mayhue, L.; Huria, H.

2012-07-01

Advanced cores and fuel assembly designs have been developed to improve operational flexibility, economic performance and further enhance safety features of nuclear power plants. The simulation of these new designs, along with strong heterogeneous fuel loading, have brought new challenges to the reactor physics methodologies currently employed in the industrial codes for core analyses. Control rod insertion during normal operation is one operational feature in the AP1000{sup R} plant of Westinghouse next generation Pressurized Water Reactor (PWR) design. This design improves its operational flexibility and efficiency but significantly challenges the conventional reactor physics methods, especially in pin power calculations. Themore » mixture loading of fuel assemblies with significant neutron spectrums causes a strong interaction between different fuel assembly types that is not fully captured with the current core design codes. To overcome the weaknesses of the conventional methods, Westinghouse has developed a state-of-the-art 3D Pin-by-Pin Calculation Methodology (P3C) and successfully implemented in the Westinghouse core design code ANC. The new methodology has been qualified and licensed for pin power prediction. The 3D P3C methodology along with its application and validation will be discussed in the paper. (authors)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, John D.; Sterbentz, James W.; Snoj, Luka

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

The concerted calculation of the BN-600 reactor for the deterministic and stochastic codes

NASA Astrophysics Data System (ADS)

Bogdanova, E. V.; Kuznetsov, A. N.

2017-01-01

The solution of the problem of increasing the safety of nuclear power plants implies the existence of complete and reliable information about the processes occurring in the core of a working reactor. Nowadays the Monte-Carlo method is the most general-purpose method used to calculate the neutron-physical characteristic of the reactor. But it is characterized by large time of calculation. Therefore, it may be useful to carry out coupled calculations with stochastic and deterministic codes. This article presents the results of research for possibility of combining stochastic and deterministic algorithms in calculation the reactor BN-600. This is only one part of the work, which was carried out in the framework of the graduation project at the NRC “Kurchatov Institute” in cooperation with S. S. Gorodkov and M. A. Kalugin. It is considering the 2-D layer of the BN-600 reactor core from the international benchmark test, published in the report IAEA-TECDOC-1623. Calculations of the reactor were performed with MCU code and then with a standard operative diffusion algorithm with constants taken from the Monte - Carlo computation. Macro cross-section, diffusion coefficients, the effective multiplication factor and the distribution of neutron flux and power were obtained in 15 energy groups. The reasonable agreement between stochastic and deterministic calculations of the BN-600 is observed.

2007 international meeting on Reduced Enrichment for Research and Test Reactors (RERTR). Abstracts and available papers presented at the meeting

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

2008-07-15

The Meeting papers discuss research and test reactor fuel performance, manufacturing and testing. Some of the main topics are: conversion from HEU to LEU in different reactors and corresponding problems and activities; flux performance and core lifetime analysis with HEU and LEU fuels; physics and safety characteristics; measurement of gamma field parameters in core with LEU fuel; nondestructive analysis of RERTR fuel; thermal hydraulic analysis; fuel interactions; transient analyses and thermal hydraulics for HEU and LEU cores; microstructure research reactor fuels; post irradiation analysis and performance; computer codes and other related problems.

Evaluation of CFETR as a Fusion Nuclear Science Facility using multiple system codes

NASA Astrophysics Data System (ADS)

Chan, V. S.; Costley, A. E.; Wan, B. N.; Garofalo, A. M.; Leuer, J. A.

2015-02-01

This paper presents the results of a multi-system codes benchmarking study of the recently published China Fusion Engineering Test Reactor (CFETR) pre-conceptual design (Wan et al 2014 IEEE Trans. Plasma Sci. 42 495). Two system codes, General Atomics System Code (GASC) and Tokamak Energy System Code (TESC), using different methodologies to arrive at CFETR performance parameters under the same CFETR constraints show that the correlation between the physics performance and the fusion performance is consistent, and the computed parameters are in good agreement. Optimization of the first wall surface for tritium breeding and the minimization of the machine size are highly compatible. Variations of the plasma currents and profiles lead to changes in the required normalized physics performance, however, they do not significantly affect the optimized size of the machine. GASC and TESC have also been used to explore a lower aspect ratio, larger volume plasma taking advantage of the engineering flexibility in the CFETR design. Assuming the ITER steady-state scenario physics, the larger plasma together with a moderately higher BT and Ip can result in a high gain Qfus ˜ 12, Pfus ˜ 1 GW machine approaching DEMO-like performance. It is concluded that the CFETR baseline mode can meet the minimum goal of the Fusion Nuclear Science Facility (FNSF) mission and advanced physics will enable it to address comprehensively the outstanding critical technology gaps on the path to a demonstration reactor (DEMO). Before proceeding with CFETR construction steady-state operation has to be demonstrated, further development is needed to solve the divertor heat load issue, and blankets have to be designed with tritium breeding ratio (TBR) >1 as a target.

Simulation of Watts Bar Unit 1 Initial Startup Tests with Continuous Energy Monte Carlo Methods

DOE Office of Scientific and Technical Information (OSTI.GOV)

Godfrey, Andrew T; Gehin, Jess C; Bekar, Kursat B

2014-01-01

The Consortium for Advanced Simulation of Light Water Reactors* is developing a collection of methods and software products known as VERA, the Virtual Environment for Reactor Applications. One component of the testing and validation plan for VERA is comparison of neutronics results to a set of continuous energy Monte Carlo solutions for a range of pressurized water reactor geometries using the SCALE component KENO-VI developed by Oak Ridge National Laboratory. Recent improvements in data, methods, and parallelism have enabled KENO, previously utilized predominately as a criticality safety code, to demonstrate excellent capability and performance for reactor physics applications. The highlymore » detailed and rigorous KENO solutions provide a reliable nu-meric reference for VERAneutronics and also demonstrate the most accurate predictions achievable by modeling and simulations tools for comparison to operating plant data. This paper demonstrates the performance of KENO-VI for the Watts Bar Unit 1 Cycle 1 zero power physics tests, including reactor criticality, control rod worths, and isothermal temperature coefficients.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Merzari, E.; Yuan, Haomin; Kraus, A.

The NEAMS program aims to develop an integrated multi-physics simulation capability “pellet-to-plant” for the design and analysis of future generations of nuclear power plants. In particular, the Reactor Product Line code suite's multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. Flow-induced vibration (FIV) is widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occurs. Given the importance of reliable componentsmore » in the nuclear industry, flow-induced vibration has long been a major concern in safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from flow-induced vibration and related failures. Advanced reactors, such as integral Pressurized Water Reactors (PWRs) considered for Small Modular Reactors (SMR), often rely on innovative component designs to meet cost and safety targets. One component that is the subject of advanced designs is the steam generator, some designs of which forego the usual shell-and-tube architecture in order to fit within the primary vessel. In addition to being more cost- and space-efficient, such steam generators need to be more reliable, since failure of the primary vessel represents a potential loss of coolant and a safety concern. A significant amount of data exists on flow-induced vibration in shell-and-tube heat exchangers, and heuristic methods are available to predict their occurrence based on a set of given assumptions. In contrast, advanced designs have far less data available. Advanced modeling and simulation based on coupled structural and fluid simulations have the potential to predict flow-induced vibration in a variety of designs, reducing the need for expensive experimental programs, especially at the design stage. Over the past five years, the Reactor Product Line has developed the integrated multi-physics code suite SHARP. The goal of developing such a tool is to perform multi-physics neutronics, thermal/fluid, and structural mechanics modeling of the components inside the full reactor core or portions of it with a user-specified fidelity. In particular SHARP contains high-fidelity single-physics codes Diablo for structural mechanics and Nek5000 for fluid mechanics calculations. Both codes are state-of-the-art, highly scalable tools that have been extensively validated. These tools form a strong basis on which to build a flow-induced vibration modeling capability. In this report we discuss one-way coupled calculations performed with Nek5000 and Diablo aimed at simulating available FIV experiments in helical steam generators in the turbulent buffeting regime. In this regime one-way coupling is judged sufficient because the pressure loads do not cause substantial displacements. It is also the most common source of vibration in helical steam generators at the low flows expected in integral PWRs. The legacy data is obtained from two datasets developed at Argonne and B&W.« less

HTR-PROTEUS pebble bed experimental program cores 9 & 10: columnar hexagonal point-on-point packing with a 1:1 moderator-to-fuel pebble ratio

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, John D.

2014-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 5, 6, 7, & 8: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:2 MODERATOR-TO-FUEL PEBBLE RATIO

DOE Office of Scientific and Technical Information (OSTI.GOV)

John D. Bess

2013-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 9 & 10: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

DOE Office of Scientific and Technical Information (OSTI.GOV)

John D. Bess

2013-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

Deterministic Modeling of the High Temperature Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ortensi, J.; Cogliati, J. J.; Pope, M. A.

2010-06-01

Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine INL’s current prismatic reactor deterministic analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 column thin annular core, and the fully loaded core critical condition with 30 columns. Special emphasis is devoted to the annular core modeling, which shares more characteristics with the NGNP base design. The DRAGON code is usedmore » in this study because it offers significant ease and versatility in modeling prismatic designs. Despite some geometric limitations, the code performs quite well compared to other lattice physics codes. DRAGON can generate transport solutions via collision probability (CP), method of characteristics (MOC), and discrete ordinates (Sn). A fine group cross section library based on the SHEM 281 energy structure is used in the DRAGON calculations. HEXPEDITE is the hexagonal z full core solver used in this study and is based on the Green’s Function solution of the transverse integrated equations. In addition, two Monte Carlo (MC) based codes, MCNP5 and PSG2/SERPENT, provide benchmarking capability for the DRAGON and the nodal diffusion solver codes. The results from this study show a consistent bias of 2–3% for the core multiplication factor. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement stems from the fact that during the experiments the control rods were adjusted to maintain criticality, whereas in the model, the rod positions were fixed. In addition, this work includes a brief study of a cross section generation approach that seeks to decouple the domain in order to account for neighbor effects. This spectral interpenetration is a dominant effect in annular HTR physics. This analysis methodology should be further explored in order to reduce the error that is systematically propagated in the traditional generation of cross sections.« less

Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murata, K.K.; Williams, D.C.; Griffith, R.O.

1997-12-01

The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of themore » input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.« less

Reactivity-worth estimates of the OSMOSE samples in the MINERVE reactor R1-MOX, R2-UO2 and MORGANE/R configurations.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhong, Z.; Klann, R. T.; Nuclear Engineering Division

2007-08-03

An initial series of calculations of the reactivity-worth of the OSMOSE samples in the MINERVE reactor with the R2-UO2 and MORGANE/R core configuration were completed. The calculation model was generated using the lattice physics code DRAGON. In addition, an initial comparison of calculated values to experimental measurements was performed based on preliminary results for the R1-MOX configuration.

Improvement of COBRA-TF for modeling of PWR cold- and hot-legs during reactor transients

NASA Astrophysics Data System (ADS)

Salko, Robert K.

COBRA-TF is a two-phase, three-field (liquid, vapor, droplets) thermal-hydraulic modeling tool that has been developed by the Pacific Northwest Laboratory under sponsorship of the NRC. The code was developed for Light Water Reactor analysis starting in the 1980s; however, its development has continued to this current time. COBRA-TF still finds wide-spread use throughout the nuclear engineering field, including nuclear-power vendors, academia, and research institutions. It has been proposed that extension of the COBRA-TF code-modeling region from vessel-only components to Pressurized Water Reactor (PWR) coolant-line regions can lead to improved Loss-of-Coolant Accident (LOCA) analysis. Improved modeling is anticipated due to COBRA-TF's capability to independently model the entrained-droplet flow-field behavior, which has been observed to impact delivery to the core region[1]. Because COBRA-TF was originally developed for vertically-dominated, in-vessel, sub-channel flow, extension of the COBRA-TF modeling region to the horizontal-pipe geometries of the coolant-lines required several code modifications, including: • Inclusion of the stratified flow regime into the COBRA-TF flow regime map, along with associated interfacial drag, wall drag and interfacial heat transfer correlations, • Inclusion of a horizontal-stratification force between adjacent mesh cells having unequal levels of stratified flow, and • Generation of a new code-input interface for the modeling of coolant-lines. The sheer number of COBRA-TF modifications that were required to complete this work turned this project into a code-development project as much as it was a study of thermal-hydraulics in reactor coolant-lines. The means for achieving these tasks shifted along the way, ultimately leading the development of a separate, nearly completely independent one-dimensional, two-phase-flow modeling code geared toward reactor coolant-line analysis. This developed code has been named CLAP, for Coolant-Line-Analysis Package. Versions were created that were both coupled to COBRA-TF and standalone, with the most recent version being a standalone code. This code performs a separate, simplified, 1-D solution of the conservation equations while making special considerations for coolant-line geometry and flow phenomena. The end of this project saw a functional code package that demonstrates a stable numerical solution and that has gone through a series of Validation and Verification tests using the Two-Phase Testing Facility (TPTF) experimental data[2]. The results indicate that CLAP is under-performing RELAP5-MOD3 in predicting the experimental void of the TPTF facility in some cases. There is no apparent pattern, however, to point to a consistent type of case that the code fails to predict properly (e.g., low-flow, high-flow, discharging to full vessel, or discharging to empty vessel). Pressure-profile predictions are sometimes unrealistic, which indicates that there may be a problem with test-case boundary conditions or with the coupling of continuity and momentum equations in the solution algorithm. The code does predict the flow regime correctly for all cases with the stratification-force model off. Turning the stratification model on can cause the low-flow case void profiles to over-react to the force and the flow regime to transition out of stratified flow. The code would benefit from an increased amount of Validation & Verification testing. The development of CLAP was significant, as it is a cleanly written, logical representation of the reactor coolant-line geometry. It is stable and capable of modeling basic flow physics in the reactor coolant-line. Code development and debugging required the temporary removal of the energy equation and mass-transfer terms in governing equations. The reintroduction of these terms will allow future coupling to RELAP and re-coupling with COBRA-TF. Adding in more applicable entrainment and de-entrainment models would allow the capture of more advanced physics in the coolant-line that can be expected during Loss-of-Coolant Accident. One of the package's benefits is its ability to be used as a platform for future coolant-line model development and implementation, including capturing of the important de-entrainment behavior in reactor hot-legs (steam-binding effect) and flow convection in the upper-plenum region of the vessel.

Status of thermalhydraulic modelling and assessment: Open issues

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bestion, D.; Barre, F.

1997-07-01

This paper presents the status of the physical modelling in present codes used for Nuclear Reactor Thermalhydraulics (TRAC, RELAP 5, CATHARE, ATHLET,...) and attempts to list the unresolved or partially resolved issues. First, the capabilities and limitations of present codes are presented. They are mainly known from a synthesis of the assessment calculations performed for both separate effect tests and integral effect tests. It is also interesting to list all the assumptions and simplifications which were made in the establishment of the system of equations and of the constitutive relations. Many of the present limitations are associated to physical situationsmore » where these assumptions are not valid. Then, recommendations are proposed to extend the capabilities of these codes.« less

Oak Ridge National Laboratory Support of Non-light Water Reactor Technologies: Capabilities Assessment for NRC Near-term Implementation Action Plans for Non-light Water Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Belles, Randy; Jain, Prashant K.; Powers, Jeffrey J.

The Oak Ridge National Laboratory (ORNL) has a rich history of support for light water reactor (LWR) and non-LWR technologies. The ORNL history involves operation of 13 reactors at ORNL including the graphite reactor dating back to World War II, two aqueous homogeneous reactors, two molten salt reactors (MSRs), a fast-burst health physics reactor, and seven LWRs. Operation of the High Flux Isotope Reactor (HFIR) has been ongoing since 1965. Expertise exists amongst the ORNL staff to provide non-LWR training; support evaluation of non-LWR licensing and safety issues; perform modeling and simulation using advanced computational tools; run laboratory experiments usingmore » equipment such as the liquid salt component test facility; and perform in-depth fuel performance and thermal-hydraulic technology reviews using a vast suite of computer codes and tools. Summaries of this expertise are included in this paper.« less

Demonstration of fully coupled simplified extended station black-out accident simulation with RELAP-7

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhao, Haihua; Zhang, Hongbin; Zou, Ling

2014-10-01

The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). The RELAP-7 code develop-ment effort started in October of 2011 and by the end of the second development year, a number of physical components with simplified two phase flow capability have been de-veloped to support the simplified boiling water reactor (BWR) extended station blackout (SBO) analyses. The demonstration case includes the major components for the primary system of a BWR, as well as the safety system components for the safety relief valve (SRV), the reactor core isolation cooling (RCIC)more » system, and the wet well. Three scenar-ios for the SBO simulations have been considered. Since RELAP-7 is not a severe acci-dent analysis code, the simulation stops when fuel clad temperature reaches damage point. Scenario I represents an extreme station blackout accident without any external cooling and cooling water injection. The system pressure is controlled by automatically releasing steam through SRVs. Scenario II includes the RCIC system but without SRV. The RCIC system is fully coupled with the reactor primary system and all the major components are dynamically simulated. The third scenario includes both the RCIC system and the SRV to provide a more realistic simulation. This paper will describe the major models and dis-cuss the results for the three scenarios. The RELAP-7 simulations for the three simplified SBO scenarios show the importance of dynamically simulating the SRVs, the RCIC sys-tem, and the wet well system to the reactor safety during extended SBO accidents.« less

Extension of the XGC code for global gyrokinetic simulations in stellarator geometry

NASA Astrophysics Data System (ADS)

Cole, Michael; Moritaka, Toseo; White, Roscoe; Hager, Robert; Ku, Seung-Hoe; Chang, Choong-Seock

2017-10-01

In this work, the total-f, gyrokinetic particle-in-cell code XGC is extended to treat stellarator geometries. Improvements to meshing tools and the code itself have enabled the first physics studies, including single particle tracing and flux surface mapping in the magnetic geometry of the heliotron LHD and quasi-isodynamic stellarator Wendelstein 7-X. These have provided the first successful test cases for our approach. XGC is uniquely placed to model the complex edge physics of stellarators. A roadmap to such a global confinement modeling capability will be presented. Single particle studies will include the physics of energetic particles' global stochastic motions and their effect on confinement. Good confinement of energetic particles is vital for a successful stellarator reactor design. These results can be compared in the core region with those of other codes, such as ORBIT3d. In subsequent work, neoclassical transport and turbulence can then be considered and compared to results from codes such as EUTERPE and GENE. After sufficient verification in the core region, XGC will move into the stellarator edge region including the material wall and neutral particle recycling.

Numerical Simulation of Measurements during the Reactor Physical Startup at Unit 3 of Rostov NPP

NASA Astrophysics Data System (ADS)

Tereshonok, V. A.; Kryakvin, L. V.; Pitilimov, V. A.; Karpov, S. A.; Kulikov, V. I.; Zhylmaganbetov, N. M.; Kavun, O. Yu.; Popykin, A. I.; Shevchenko, R. A.; Shevchenko, S. A.; Semenova, T. V.

2017-12-01

The results of numerical calculations and measurements of some reactor parameters during the physical startup tests at unit 3 of Rostov NPP are presented. The following parameters are considered: the critical boron acid concentration and the currents from ionization chambers (IC) during the scram system efficiency evaluation. The scram system efficiency was determined using the inverse point kinetics equation with the measured and simulated IC currents. The results of steady-state calculations of relative power distribution and efficiency of the scram system and separate groups of control rods of the control and protection system are also presented. The calculations are performed using several codes, including precision ones.

Development and Implementation of CFD-Informed Models for the Advanced Subchannel Code CTF

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blyth, Taylor S.; Avramova, Maria

The research described in this PhD thesis contributes to the development of efficient methods for utilization of high-fidelity models and codes to inform low-fidelity models and codes in the area of nuclear reactor core thermal-hydraulics. The objective is to increase the accuracy of predictions of quantities of interests using high-fidelity CFD models while preserving the efficiency of low-fidelity subchannel core calculations. An original methodology named Physics- based Approach for High-to-Low Model Information has been further developed and tested. The overall physical phenomena and corresponding localized effects, which are introduced by the presence of spacer grids in light water reactor (LWR)more » cores, are dissected in corresponding four building basic processes, and corresponding models are informed using high-fidelity CFD codes. These models are a spacer grid-directed cross-flow model, a grid-enhanced turbulent mixing model, a heat transfer enhancement model, and a spacer grid pressure loss model. The localized CFD-models are developed and tested using the CFD code STAR-CCM+, and the corresponding global model development and testing in sub-channel formulation is performed in the thermal- hydraulic subchannel code CTF. The improved CTF simulations utilize data-files derived from CFD STAR-CCM+ simulation results covering the spacer grid design desired for inclusion in the CTF calculation. The current implementation of these models is examined and possibilities for improvement and further development are suggested. The validation experimental database is extended by including the OECD/NRC PSBT benchmark data. The outcome is an enhanced accuracy of CTF predictions while preserving the computational efficiency of a low-fidelity subchannel code.« less

Development and Implementation of CFD-Informed Models for the Advanced Subchannel Code CTF

NASA Astrophysics Data System (ADS)

Blyth, Taylor S.

The research described in this PhD thesis contributes to the development of efficient methods for utilization of high-fidelity models and codes to inform low-fidelity models and codes in the area of nuclear reactor core thermal-hydraulics. The objective is to increase the accuracy of predictions of quantities of interests using high-fidelity CFD models while preserving the efficiency of low-fidelity subchannel core calculations. An original methodology named Physics-based Approach for High-to-Low Model Information has been further developed and tested. The overall physical phenomena and corresponding localized effects, which are introduced by the presence of spacer grids in light water reactor (LWR) cores, are dissected in corresponding four building basic processes, and corresponding models are informed using high-fidelity CFD codes. These models are a spacer grid-directed cross-flow model, a grid-enhanced turbulent mixing model, a heat transfer enhancement model, and a spacer grid pressure loss model. The localized CFD-models are developed and tested using the CFD code STAR-CCM+, and the corresponding global model development and testing in sub-channel formulation is performed in the thermal-hydraulic subchannel code CTF. The improved CTF simulations utilize data-files derived from CFD STAR-CCM+ simulation results covering the spacer grid design desired for inclusion in the CTF calculation. The current implementation of these models is examined and possibilities for improvement and further development are suggested. The validation experimental database is extended by including the OECD/NRC PSBT benchmark data. The outcome is an enhanced accuracy of CTF predictions while preserving the computational efficiency of a low-fidelity subchannel code.

Development of high-fidelity multiphysics system for light water reactor analysis

NASA Astrophysics Data System (ADS)

Magedanz, Jeffrey W.

There has been a tendency in recent years toward greater heterogeneity in reactor cores, due to the use of mixed-oxide (MOX) fuel, burnable absorbers, and longer cycles with consequently higher fuel burnup. The resulting asymmetry of the neutron flux and energy spectrum between regions with different compositions causes a need to account for the directional dependence of the neutron flux, instead of the traditional diffusion approximation. Furthermore, the presence of both MOX and high-burnup fuel in the core increases the complexity of the heat conduction. The heat transfer properties of the fuel pellet change with irradiation, and the thermal and mechanical expansion of the pellet and cladding strongly affect the size of the gap between them, and its consequent thermal resistance. These operational tendencies require higher fidelity multi-physics modeling capabilities, and this need is addressed by the developments performed within this PhD research. The dissertation describes the development of a High-Fidelity Multi-Physics System for Light Water Reactor Analysis. It consists of three coupled codes -- CTF for Thermal Hydraulics, TORT-TD for Neutron Kinetics, and FRAPTRAN for Fuel Performance. It is meant to address these modeling challenges in three ways: (1) by resolving the state of the system at the level of each fuel pin, rather than homogenizing entire fuel assemblies, (2) by using the multi-group Discrete Ordinates method to account for the directional dependence of the neutron flux, and (3) by using a fuel-performance code, rather than a Thermal Hydraulics code's simplified fuel model, to account for the material behavior of the fuel and its feedback to the hydraulic and neutronic behavior of the system. While the first two are improvements, the third, the use of a fuel-performance code for feedback, constitutes an innovation in this PhD project. Also important to this work is the manner in which such coupling is written. While coupling involves combining codes into a single executable, they are usually still developed and maintained separately. It should thus be a design objective to minimize the changes to those codes, and keep the changes to each code free of dependence on the details of the other codes. This will ease the incorporation of new versions of the code into the coupling, as well as re-use of parts of the coupling to couple with different codes. In order to fulfill this objective, an interface for each code was created in the form of an object-oriented abstract data type. Object-oriented programming is an effective method for enforcing a separation between different parts of a program, and clarifying the communication between them. The interfaces enable the main program to control the codes in terms of high-level functionality. This differs from the established practice of a master/slave relationship, in which the slave code is incorporated into the master code as a set of subroutines. While this PhD research continues previous work with a coupling between CTF and TORT-TD, it makes two major original contributions: (1) using a fuel-performance code, instead of a thermal-hydraulics code's simplified built-in models, to model the feedback from the fuel rods, and (2) the design of an object-oriented interface as an innovative method to interact with a coupled code in a high-level, easily-understandable manner. The resulting code system will serve as a tool to study the question of under what conditions, and to what extent, these higher-fidelity methods will provide benefits to reactor core analysis. (Abstract shortened by UMI.)

Code qualification of structural materials for AFCI advanced recycling reactors.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Natesan, K.; Li, M.; Majumdar, S.

2012-05-31

This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Codemore » Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP) and the Power Reactor Innovative Small Module (PRISM), the NRC/Advisory Committee on Reactor Safeguards (ACRS) raised numerous safety-related issues regarding elevated-temperature structural integrity criteria. Most of these issues remained unresolved today. These critical licensing reviews provide a basis for the evaluation of underlying technical issues for future advanced sodium-cooled reactors. Major materials performance issues and high temperature design methodology issues pertinent to the ARR are addressed in the report. The report is organized as follows: the ARR reference design concepts proposed by the Argonne National Laboratory and four industrial consortia were reviewed first, followed by a summary of the major code qualification and licensing issues for the ARR structural materials. The available database is presented for the ASME Code-qualified structural alloys (e.g. 304, 316 stainless steels, 2.25Cr-1Mo, and mod.9Cr-1Mo), including physical properties, tensile properties, impact properties and fracture toughness, creep, fatigue, creep-fatigue interaction, microstructural stability during long-term thermal aging, material degradation in sodium environments and effects of neutron irradiation for both base metals and weld metals. An assessment of modified versions of Type 316 SS, i.e. Type 316LN and its Japanese version, 316FR, was conducted to provide a perspective for codification of 316LN or 316FR in Subsection NH. Current status and data availability of four new advanced alloys, i.e. NF616, NF616+TMT, NF709, and HT-UPS, are also addressed to identify the R&D needs for their code qualification for ARR applications. For both conventional and new alloys, issues related to high temperature design methodology are described to address the needs for improvements for the ARR design and licensing. Assessments have shown that there are significant data gaps for the full qualification and licensing of the ARR structural materials. Development and evaluation of structural materials require a variety of experimental facilities that have been seriously degraded in the past. The availability and additional needs for the key experimental facilities are summarized at the end of the report. Detailed information covered in each Chapter is given.« less

Modeling of a Flooding Induced Station Blackout for a Pressurized Water Reactor Using the RISMC Toolkit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mandelli, Diego; Prescott, Steven R; Smith, Curtis L

2011-07-01

In the Risk Informed Safety Margin Characterization (RISMC) approach we want to understand not just the frequency of an event like core damage, but how close we are (or are not) to key safety-related events and how might we increase our safety margins. The RISMC Pathway uses the probabilistic margin approach to quantify impacts to reliability and safety by coupling both probabilistic (via stochastic simulation) and mechanistic (via physics models) approaches. This coupling takes place through the interchange of physical parameters and operational or accident scenarios. In this paper we apply the RISMC approach to evaluate the impact of amore » power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., system activation) and to perform statistical analyses (e.g., run multiple RELAP-7 simulations where sequencing/timing of events have been changed according to a set of stochastic distributions). By using the RISMC toolkit, we can evaluate how power uprate affects the system recovery measures needed to avoid core damage after the PWR lost all available AC power by a tsunami induced flooding. The simulation of the actual flooding is performed by using a smooth particle hydrodynamics code: NEUTRINO.« less

Heuristic rules embedded genetic algorithm for in-core fuel management optimization

NASA Astrophysics Data System (ADS)

Alim, Fatih

The objective of this study was to develop a unique methodology and a practical tool for designing loading pattern (LP) and burnable poison (BP) pattern for a given Pressurized Water Reactor (PWR) core. Because of the large number of possible combinations for the fuel assembly (FA) loading in the core, the design of the core configuration is a complex optimization problem. It requires finding an optimal FA arrangement and BP placement in order to achieve maximum cycle length while satisfying the safety constraints. Genetic Algorithms (GA) have been already used to solve this problem for LP optimization for both PWR and Boiling Water Reactor (BWR). The GA, which is a stochastic method works with a group of solutions and uses random variables to make decisions. Based on the theories of evaluation, the GA involves natural selection and reproduction of the individuals in the population for the next generation. The GA works by creating an initial population, evaluating it, and then improving the population by using the evaluation operators. To solve this optimization problem, a LP optimization package, GARCO (Genetic Algorithm Reactor Code Optimization) code is developed in the framework of this thesis. This code is applicable for all types of PWR cores having different geometries and structures with an unlimited number of FA types in the inventory. To reach this goal, an innovative GA is developed by modifying the classical representation of the genotype. To obtain the best result in a shorter time, not only the representation is changed but also the algorithm is changed to use in-core fuel management heuristics rules. The improved GA code was tested to demonstrate and verify the advantages of the new enhancements. The developed methodology is explained in this thesis and preliminary results are shown for the VVER-1000 reactor hexagonal geometry core and the TMI-1 PWR. The improved GA code was tested to verify the advantages of new enhancements. The core physics code used for VVER in this research is Moby-Dick, which was developed to analyze the VVER by SKODA Inc. The SIMULATE-3 code, which is an advanced two-group nodal code, is used to analyze the TMI-1.

Fuel burnup analysis for IRIS reactor using MCNPX and WIMS-D5 codes

NASA Astrophysics Data System (ADS)

Amin, E. A.; Bashter, I. I.; Hassan, Nabil M.; Mustafa, S. S.

2017-02-01

International Reactor Innovative and Secure (IRIS) reactor is a compact power reactor designed with especial features. It contains Integral Fuel Burnable Absorber (IFBA). The core is heterogeneous both axially and radially. This work provides the full core burn up analysis for IRIS reactor using MCNPX and WIMDS-D5 codes. Criticality calculations, radial and axial power distributions and nuclear peaking factor at the different stages of burnup were studied. Effective multiplication factor values for the core were estimated by coupling MCNPX code with WIMS-D5 code and compared with SAS2H/KENO-V code values at different stages of burnup. The two calculation codes show good agreement and correlation. The values of radial and axial powers for the full core were also compared with published results given by SAS2H/KENO-V code (at the beginning and end of reactor operation). The behavior of both radial and axial power distribution is quiet similar to the other data published by SAS2H/KENO-V code. The peaking factor values estimated in the present work are close to its values calculated by SAS2H/KENO-V code.

Evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Tikhomirov, Georgy; Ternovykh, Mikhail; Saldikov, Ivan; Fomichenko, Peter; Gerasimov, Alexander

2017-09-01

The strategy of the development of nuclear power in Russia provides for use of fast power reactors in closed nuclear fuel cycle. The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of energy. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. The closed nuclear fuel cycle concept of the PRORYV assumes self-supplied mode of operation with fuel regeneration by neutron capture reaction in non-enriched uranium, which is used as a raw material. Operating modes of reactors and its characteristics should be chosen so as to provide the self-sufficient mode by using of fissile isotopes while refueling by depleted uranium and to support this state during the entire period of reactor operation. Thus, the actual issue is modeling fuel handling processes. To solve these problems, the code REPRORYV (Recycle for PRORYV) has been developed. It simulates nuclide streams in non-reactor stages of the closed fuel cycle. At the same time various verified codes can be used to evaluate in-core characteristics of a reactor. By using this approach various options for nuclide streams and assess the impact of different plutonium content in the fuel, fuel processing conditions, losses during fuel processing, as well as the impact of initial uncertainties on neutron-physical characteristics of reactor are considered in this study.

Structural mechanics simulations

NASA Technical Reports Server (NTRS)

Biffle, Johnny H.

1992-01-01

Sandia National Laboratory has a very broad structural capability. Work has been performed in support of reentry vehicles, nuclear reactor safety, weapons systems and components, nuclear waste transport, strategic petroleum reserve, nuclear waste storage, wind and solar energy, drilling technology, and submarine programs. The analysis environment contains both commercial and internally developed software. Included are mesh generation capabilities, structural simulation codes, and visual codes for examining simulation results. To effectively simulate a wide variety of physical phenomena, a large number of constitutive models have been developed.

The Serpent Monte Carlo Code: Status, Development and Applications in 2013

NASA Astrophysics Data System (ADS)

Leppänen, Jaakko; Pusa, Maria; Viitanen, Tuomas; Valtavirta, Ville; Kaltiaisenaho, Toni

2014-06-01

The Serpent Monte Carlo reactor physics burnup calculation code has been developed at VTT Technical Research Centre of Finland since 2004, and is currently used in 100 universities and research organizations around the world. This paper presents the brief history of the project, together with the currently available methods and capabilities and plans for future work. Typical user applications are introduced in the form of a summary review on Serpent-related publications over the past few years.

PROTEUS-SN User Manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shemon, Emily R.; Smith, Micheal A.; Lee, Changho

2016-02-16

PROTEUS-SN is a three-dimensional, highly scalable, high-fidelity neutron transport code developed at Argonne National Laboratory. The code is applicable to all spectrum reactor transport calculations, particularly those in which a high degree of fidelity is needed either to represent spatial detail or to resolve solution gradients. PROTEUS-SN solves the second order formulation of the transport equation using the continuous Galerkin finite element method in space, the discrete ordinates approximation in angle, and the multigroup approximation in energy. PROTEUS-SN’s parallel methodology permits the efficient decomposition of the problem by both space and angle, permitting large problems to run efficiently on hundredsmore » of thousands of cores. PROTEUS-SN can also be used in serial or on smaller compute clusters (10’s to 100’s of cores) for smaller homogenized problems, although it is generally more computationally expensive than traditional homogenized methodology codes. PROTEUS-SN has been used to model partially homogenized systems, where regions of interest are represented explicitly and other regions are homogenized to reduce the problem size and required computational resources. PROTEUS-SN solves forward and adjoint eigenvalue problems and permits both neutron upscattering and downscattering. An adiabatic kinetics option has recently been included for performing simple time-dependent calculations in addition to standard steady state calculations. PROTEUS-SN handles void and reflective boundary conditions. Multigroup cross sections can be generated externally using the MC2-3 fast reactor multigroup cross section generation code or internally using the cross section application programming interface (API) which can treat the subgroup or resonance table libraries. PROTEUS-SN is written in Fortran 90 and also includes C preprocessor definitions. The code links against the PETSc, METIS, HDF5, and MPICH libraries. It optionally links against the MOAB library and is a part of the SHARP multi-physics suite for coupled multi-physics analysis of nuclear reactors. This user manual describes how to set up a neutron transport simulation with the PROTEUS-SN code. A companion methodology manual describes the theory and algorithms within PROTEUS-SN.« less

Monte Carlo simulation of ion-material interactions in nuclear fusion devices

NASA Astrophysics Data System (ADS)

Nieto Perez, M.; Avalos-Zuñiga, R.; Ramos, G.

2017-06-01

One of the key aspects regarding the technological development of nuclear fusion reactors is the understanding of the interaction between high-energy ions coming from the confined plasma and the materials that the plasma-facing components are made of. Among the multiple issues important to plasma-wall interactions in fusion devices, physical erosion and composition changes induced by energetic particle bombardment are considered critical due to possible material flaking, changes to surface roughness, impurity transport and the alteration of physicochemical properties of the near surface region due to phenomena such as redeposition or implantation. A Monte Carlo code named MATILDA (Modeling of Atomic Transport in Layered Dynamic Arrays) has been developed over the years to study phenomena related to ion beam bombardment such as erosion rate, composition changes, interphase mixing and material redeposition, which are relevant issues to plasma-aided manufacturing of microelectronics, components on object exposed to intense solar wind, fusion reactor technology and other important industrial fields. In the present work, the code is applied to study three cases of plasma material interactions relevant to fusion devices in order to highlight the code's capabilities: (1) the Be redeposition process on the ITER divertor, (2) physical erosion enhancement in castellated surfaces and (3) damage to multilayer mirrors used on EUV diagnostics in fusion devices due to particle bombardment.

The scheme for evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Saldikov, I. S.; Ternovykh, M. Yu; Fomichenko, P. A.; Gerasimov, A. S.

2017-01-01

The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of power. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. To solve the closed nuclear fuel modeling tasks REPRORYV code was developed. It simulates the mass flow for nuclides in the closed fuel cycle. This paper presents the results of modeling of a closed nuclear fuel cycle, nuclide flows considering the influence of the uncertainty on the outcome of neutron-physical characteristics of the reactor.

GAMSOR: Gamma Source Preparation and DIF3D Flux Solution

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, M. A.; Lee, C. H.; Hill, R. N.

2016-12-15

Nuclear reactors that rely upon the fission reaction have two modes of thermal energy deposition in the reactor system: neutron absorption and gamma absorption. The gamma rays are typically generated by neutron absorption reactions or during the fission process which means the primary driver of energy production is of course the neutron interaction. In conventional reactor physics methods, the gamma heating component is ignored such that the gamma absorption is forced to occur at the gamma emission site. For experimental reactor systems like EBR-II and FFTF, the placement of structural pins and assemblies internal to the core leads to problemsmore » with power heating predictions because there is no fission power source internal to the assembly to dictate a spatial distribution of the power. As part of the EBR-II support work in the 1980s, the GAMSOR code was developed to assist analysts in calculating the gamma heating. The GAMSOR code is a modified version of DIF3D and actually functions within a sequence of DIF3D calculations. The gamma flux in a conventional fission reactor system does not perturb the neutron flux and thus the gamma flux calculation can be cast as a fixed source problem given a solution to the steady state neutron flux equation. This leads to a sequence of DIF3D calculations, called the GAMSOR sequence, which involves solving the neutron flux, then the gamma flux, then combining the results to do a summary edit. In this manuscript, we go over the GAMSOR code and detail how it is put together and functions. We also discuss how to setup the GAMSOR sequence and input for each DIF3D calculation in the GAMSOR sequence. With the GAMSOR capability, users can take any valid steady state DIF3D calculation and compute the power distribution due to neutron and gamma heating. The MC2-3 code is the preferable companion code to use for generating neutron and gamma cross section data, but the GAMSOR code can accept cross section data from other sources. To further this aspect, an additional utility code was created which demonstrates how to merge the neutron and gamma cross section data together to carry out a simultaneous solve of the two systems.« less

Development of a Model and Computer Code to Describe Solar Grade Silicon Production Processes

NASA Technical Reports Server (NTRS)

Srivastava, R.; Gould, R. K.

1979-01-01

Mathematical models and computer codes based on these models, which allow prediction of the product distribution in chemical reactors for converting gaseous silicon compounds to condensed-phase silicon were developed. The following tasks were accomplished: (1) formulation of a model for silicon vapor separation/collection from the developing turbulent flow stream within reactors of the Westinghouse (2) modification of an available general parabolic code to achieve solutions to the governing partial differential equations (boundary layer type) which describe migration of the vapor to the reactor walls, (3) a parametric study using the boundary layer code to optimize the performance characteristics of the Westinghouse reactor, (4) calculations relating to the collection efficiency of the new AeroChem reactor, and (5) final testing of the modified LAPP code for use as a method of predicting Si(1) droplet sizes in these reactors.

Uncertainty analysis on reactivity and discharged inventory for a pressurized water reactor fuel assembly due to {sup 235,238}U nuclear data uncertainties

DOE Office of Scientific and Technical Information (OSTI.GOV)

Da Cruz, D. F.; Rochman, D.; Koning, A. J.

2012-07-01

This paper discusses the uncertainty analysis on reactivity and inventory for a typical PWR fuel element as a result of uncertainties in {sup 235,238}U nuclear data. A typical Westinghouse 3-loop fuel assembly fuelled with UO{sub 2} fuel with 4.8% enrichment has been selected. The Total Monte-Carlo method has been applied using the deterministic transport code DRAGON. This code allows the generation of the few-groups nuclear data libraries by directly using data contained in the nuclear data evaluation files. The nuclear data used in this study is from the JEFF3.1 evaluation, and the nuclear data files for {sup 238}U and {supmore » 235}U (randomized for the generation of the various DRAGON libraries) are taken from the nuclear data library TENDL. The total uncertainty (obtained by randomizing all {sup 238}U and {sup 235}U nuclear data in the ENDF files) on the reactor parameters has been split into different components (different nuclear reaction channels). Results show that the TMC method in combination with a deterministic transport code constitutes a powerful tool for performing uncertainty and sensitivity analysis of reactor physics parameters. (authors)« less

Varied applications of a new maximum-likelihood code with complete covariance capability. [FERRET, for data adjustment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schmittroth, F.

1978-01-01

Applications of a new data-adjustment code are given. The method is based on a maximum-likelihood extension of generalized least-squares methods that allow complete covariance descriptions for the input data and the final adjusted data evaluations. The maximum-likelihood approach is used with a generalized log-normal distribution that provides a way to treat problems with large uncertainties and that circumvents the problem of negative values that can occur for physically positive quantities. The computer code, FERRET, is written to enable the user to apply it to a large variety of problems by modifying only the input subroutine. The following applications are discussed:more » A 75-group a priori damage function is adjusted by as much as a factor of two by use of 14 integral measurements in different reactor spectra. Reactor spectra and dosimeter cross sections are simultaneously adjusted on the basis of both integral measurements and experimental proton-recoil spectra. The simultaneous use of measured reaction rates, measured worths, microscopic measurements, and theoretical models are used to evaluate dosimeter and fission-product cross sections. Applications in the data reduction of neutron cross section measurements and in the evaluation of reactor after-heat are also considered. 6 figures.« less

Computer model of catalytic combustion/Stirling engine heater head

NASA Technical Reports Server (NTRS)

Chu, E. K.; Chang, R. L.; Tong, H.

1981-01-01

The basic Acurex HET code was modified to analyze specific problems for Stirling engine heater head applications. Specifically, the code can model: an adiabatic catalytic monolith reactor, an externally cooled catalytic cylindrical reactor/flat plate reactor, a coannular tube radiatively cooled reactor, and a monolithic reactor radiating to upstream and downstream heat exchangers.

GAMSOR: Gamma Source Preparation and DIF3D Flux Solution

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, M. A.; Lee, C. H.; Hill, R. N.

2017-06-28

Nuclear reactors that rely upon the fission reaction have two modes of thermal energy deposition in the reactor system: neutron absorption and gamma absorption. The gamma rays are typically generated by neutron capture reactions or during the fission process which means the primary driver of energy production is of course the neutron interaction. In conventional reactor physics methods, the gamma heating component is ignored such that the gamma absorption is forced to occur at the gamma emission site. For experimental reactor systems like EBR-II and FFTF, the placement of structural pins and assemblies internal to the core leads to problemsmore » with power heating predictions because there is no fission power source internal to the assembly to dictate a spatial distribution of the power. As part of the EBR-II support work in the 1980s, the GAMSOR code was developed to assist analysts in calculating the gamma heating. The GAMSOR code is a modified version of DIF3D and actually functions within a sequence of DIF3D calculations. The gamma flux in a conventional fission reactor system does not perturb the neutron flux and thus the gamma flux calculation can be cast as a fixed source problem given a solution to the steady state neutron flux equation. This leads to a sequence of DIF3D calculations, called the GAMSOR sequence, which involves solving the neutron flux, then the gamma flux, and then combining the results to do a summary edit. In this manuscript, we go over the GAMSOR code and detail how it is put together and functions. We also discuss how to setup the GAMSOR sequence and input for each DIF3D calculation in the GAMSOR sequence.« less

SHARP User Manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yu, Y. Q.; Shemon, E. R.; Thomas, J. W.

SHARP is an advanced modeling and simulation toolkit for the analysis of nuclear reactors. It is comprised of several components including physical modeling tools, tools to integrate the physics codes for multi-physics analyses, and a set of tools to couple the codes within the MOAB framework. Physics modules currently include the neutronics code PROTEUS, the thermal-hydraulics code Nek5000, and the structural mechanics code Diablo. This manual focuses on performing multi-physics calculations with the SHARP ToolKit. Manuals for the three individual physics modules are available with the SHARP distribution to help the user to either carry out the primary multi-physics calculationmore » with basic knowledge or perform further advanced development with in-depth knowledge of these codes. This manual provides step-by-step instructions on employing SHARP, including how to download and install the code, how to build the drivers for a test case, how to perform a calculation and how to visualize the results. Since SHARP has some specific library and environment dependencies, it is highly recommended that the user read this manual prior to installing SHARP. Verification tests cases are included to check proper installation of each module. It is suggested that the new user should first follow the step-by-step instructions provided for a test problem in this manual to understand the basic procedure of using SHARP before using SHARP for his/her own analysis. Both reference output and scripts are provided along with the test cases in order to verify correct installation and execution of the SHARP package. At the end of this manual, detailed instructions are provided on how to create a new test case so that user can perform novel multi-physics calculations with SHARP. Frequently asked questions are listed at the end of this manual to help the user to troubleshoot issues.« less

Approach to ignition of tokamak reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sigmar, D.J.

1981-02-01

Recent transport modeling results for JET, INTOR, and ETF are reviewed and analyzed with respect to existing uncertainties in the underlying physics, the self-consistency of the very large numerical codes, and the margin for ignition. The codes show ignition to occur in ETF/INTOR-sized machines if empirical scaling can be extrapolated to ion temperatures (and beta values) much higher than those presently achieved, if there is no significant impurity accumulation over the first 7 s, and if the known ideal and resistive MHD instabilities remain controllable for the evolving plasma profiles during ignition startup.

Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mertyurek, Ugur; Gauld, Ian C.

In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup.more » The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.« less

Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

DOE PAGES

Mertyurek, Ugur; Gauld, Ian C.

2015-12-24

In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup.more » The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.« less

Sustainable Thorium Nuclear Fuel Cycles: A Comparison of Intermediate and Fast Neutron Spectrum Systems

DOE PAGES

Brown, Nicholas R.; Powers, Jeffrey J.; Feng, B.; ...

2015-05-21

This paper presents analyses of possible reactor representations of a nuclear fuel cycle with continuous recycling of thorium and produced uranium (mostly U-233) with thorium-only feed. The analysis was performed in the context of a U.S. Department of Energy effort to develop a compendium of informative nuclear fuel cycle performance data. The objective of this paper is to determine whether intermediate spectrum systems, having a majority of fission events occurring with incident neutron energies between 1 eV and 10 5 eV, perform as well as fast spectrum systems in this fuel cycle. The intermediate spectrum options analyzed include tight latticemore » heavy or light water-cooled reactors, continuously refueled molten salt reactors, and a sodium-cooled reactor with hydride fuel. All options were modeled in reactor physics codes to calculate their lattice physics, spectrum characteristics, and fuel compositions over time. Based on these results, detailed metrics were calculated to compare the fuel cycle performance. These metrics include waste management and resource utilization, and are binned to accommodate uncertainties. The performance of the intermediate systems for this selfsustaining thorium fuel cycle was similar to a representative fast spectrum system. However, the number of fission neutrons emitted per neutron absorbed limits performance in intermediate spectrum systems.« less

Fast particles in a steady-state compact FNS and compact ST reactor

NASA Astrophysics Data System (ADS)

Gryaznevich, M. P.; Nicolai, A.; Buxton, P.

2014-10-01

This paper presents results of studies of fast particles (ions and alpha particles) in a steady-state compact fusion neutron source (CFNS) and a compact spherical tokamak (ST) reactor with Monte-Carlo and Fokker-Planck codes. Full-orbit simulations of fast particle physics indicate that a compact high field ST can be optimized for energy production by a reduction of the necessary (for the alpha containment) plasma current compared with predictions made using simple analytic expressions, or using guiding centre approximation in a numerical code. Alpha particle losses may result in significant heating and erosion of the first wall, so such losses for an ST pilot plant have been calculated and total and peak wall loads dependence on the plasma current has been studied. The problem of dilution has been investigated and results for compact and big size devices are compared.

LWR pressure vessel surveillance dosimetry improvement program: LWR power reactor surveillance physics-dosimetry data base compendium

DOE Office of Scientific and Technical Information (OSTI.GOV)

McElroy, W.N.

1985-08-01

This NRC physics-dosimetry compendium is a collation of information and data developed from available research and commercial light water reactor vessel surveillance program (RVSP) documents and related surveillance capsule reports. The data represents the results of the HEDL least-squares FERRET-SAND II Code re-evaluation of exposure units and values for 47 PWR and BWR surveillance capsules for W, B and W, CE, and GE power plants. Using a consistent set of auxiliary data and dosimetry-adjusted reactor physics results, the revised fluence values for E > 1 MeV averaged 25% higher than the originally reported values. The range of fluence values (new/old)more » was from a low of 0.80 to a high of 2.38. These HEDL-derived FERRET-SAND II exposure parameter values are being used for NRC-supported HEDL and other PWR and BWR trend curve data development and testing studies. These studies are providing results to support Revision 2 of Regulatory Guide 1.99. As stated by Randall (Ra84), the Guide is being updated to reflect recent studies of the physical basis for neutron radiation damage and efforts to correlate damage to chemical composition and fluence.« less

Development of a model and computer code to describe solar grade silicon production processes

NASA Technical Reports Server (NTRS)

Gould, R. K.; Srivastava, R.

1979-01-01

Two computer codes were developed for describing flow reactors in which high purity, solar grade silicon is produced via reduction of gaseous silicon halides. The first is the CHEMPART code, an axisymmetric, marching code which treats two phase flows with models describing detailed gas-phase chemical kinetics, particle formation, and particle growth. It can be used to described flow reactors in which reactants, mix, react, and form a particulate phase. Detailed radial gas-phase composition, temperature, velocity, and particle size distribution profiles are computed. Also, deposition of heat, momentum, and mass (either particulate or vapor) on reactor walls is described. The second code is a modified version of the GENMIX boundary layer code which is used to compute rates of heat, momentum, and mass transfer to the reactor walls. This code lacks the detailed chemical kinetics and particle handling features of the CHEMPART code but has the virtue of running much more rapidly than CHEMPART, while treating the phenomena occurring in the boundary layer in more detail.

Monte Carlo modelling of TRIGA research reactor

NASA Astrophysics Data System (ADS)

El Bakkari, B.; Nacir, B.; El Bardouni, T.; El Younoussi, C.; Merroun, O.; Htet, A.; Boulaich, Y.; Zoubair, M.; Boukhal, H.; Chakir, M.

2010-10-01

The Moroccan 2 MW TRIGA MARK II research reactor at Centre des Etudes Nucléaires de la Maâmora (CENM) achieved initial criticality on May 2, 2007. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes for their use in agriculture, industry, and medicine. This study deals with the neutronic analysis of the 2-MW TRIGA MARK II research reactor at CENM and validation of the results by comparisons with the experimental, operational, and available final safety analysis report (FSAR) values. The study was prepared in collaboration between the Laboratory of Radiation and Nuclear Systems (ERSN-LMR) from Faculty of Sciences of Tetuan (Morocco) and CENM. The 3-D continuous energy Monte Carlo code MCNP (version 5) was used to develop a versatile and accurate full model of the TRIGA core. The model represents in detailed all components of the core with literally no physical approximation. Continuous energy cross-section data from the more recent nuclear data evaluations (ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1, and JENDL-3.3) as well as S( α, β) thermal neutron scattering functions distributed with the MCNP code were used. The cross-section libraries were generated by using the NJOY99 system updated to its more recent patch file "up259". The consistency and accuracy of both the Monte Carlo simulation and neutron transport physics were established by benchmarking the TRIGA experiments. Core excess reactivity, total and integral control rods worth as well as power peaking factors were used in the validation process. Results of calculations are analysed and discussed.

NARMER-1: a photon point-kernel code with build-up factors

NASA Astrophysics Data System (ADS)

Visonneau, Thierry; Pangault, Laurence; Malouch, Fadhel; Malvagi, Fausto; Dolci, Florence

2017-09-01

This paper presents an overview of NARMER-1, the new generation of photon point-kernel code developed by the Reactor Studies and Applied Mathematics Unit (SERMA) at CEA Saclay Center. After a short introduction giving some history points and the current context of development of the code, the paper exposes the principles implemented in the calculation, the physical quantities computed and surveys the generic features: programming language, computer platforms, geometry package, sources description, etc. Moreover, specific and recent features are also detailed: exclusion sphere, tetrahedral meshes, parallel operations. Then some points about verification and validation are presented. Finally we present some tools that can help the user for operations like visualization and pre-treatment.

Neutronic calculation of fast reactors by the EUCLID/V1 integrated code

NASA Astrophysics Data System (ADS)

Koltashev, D. A.; Stakhanova, A. A.

2017-01-01

This article considers neutronic calculation of a fast-neutron lead-cooled reactor BREST-OD-300 by the EUCLID/V1 integrated code. The main goal of development and application of integrated codes is a nuclear power plant safety justification. EUCLID/V1 is integrated code designed for coupled neutronics, thermomechanical and thermohydraulic fast reactor calculations under normal and abnormal operating conditions. EUCLID/V1 code is being developed in the Nuclear Safety Institute of the Russian Academy of Sciences. The integrated code has a modular structure and consists of three main modules: thermohydraulic module HYDRA-IBRAE/LM/V1, thermomechanical module BERKUT and neutronic module DN3D. In addition, the integrated code includes databases with fuel, coolant and structural materials properties. Neutronic module DN3D provides full-scale simulation of neutronic processes in fast reactors. Heat sources distribution, control rods movement, reactivity level changes and other processes can be simulated. Neutron transport equation in multigroup diffusion approximation is solved. This paper contains some calculations implemented as a part of EUCLID/V1 code validation. A fast-neutron lead-cooled reactor BREST-OD-300 transient simulation (fuel assembly floating, decompression of passive feedback system channel) and cross-validation with MCU-FR code results are presented in this paper. The calculations demonstrate EUCLID/V1 code application for BREST-OD-300 simulating and safety justification.

Analytical methods in the high conversion reactor core design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zeggel, W.; Oldekop, W.; Axmann, J.K.

High conversion reactor (HCR) design methods have been used at the Technical University of Braunschweig (TUBS) with the technological support of Kraftwerk Union (KWU). The present state and objectives of this cooperation between KWU and TUBS in the field of HCRs have been described using existing design models and current activities aimed at further development and validation of the codes. The hard physical and thermal-hydraulic boundary conditions of pressurized water reactor (PWR) cores with a high degree of fuel utilization result from the tight packing of the HCR fuel rods and the high fissionable plutonium content of the fuel. Inmore » terms of design, the problem will be solved with rod bundles whose fuel rods are adjusted by helical spacers to the proposed small rod pitches. These HCR properties require novel computational models for neutron physics, thermal hydraulics, and fuel rod design. By means of a survey of the codes, the analytical procedure for present-day HCR core design is presented. The design programs are currently under intensive development, as design tools with a solid, scientific foundation and with essential parameters that are widely valid and are required for a promising optimization of the HCR core. Design results and a survey of future HCR development are given. In this connection, the reoptimization of the PWR core in the direction of an HCR is considered a fascinating scientific task, with respect to both economic and safety aspects.« less

Initial verification and validation of RAZORBACK - A research reactor transient analysis code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Talley, Darren G.

2015-09-01

This report describes the work and results of the initial verification and validation (V&V) of the beta release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This initial V&V effort was intended to confirm that the code work to-date shows good agreement between simulation and actualmore » ACRR operations, indicating that the subsequent V&V effort for the official release of the code will be successful.« less

Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

NASA Astrophysics Data System (ADS)

Grodzki, Marcin; Darnowski, Piotr; Niewiński, Grzegorz

2017-12-01

The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an `early design' variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit). A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

Relaunch of the Interactive Plasma Physics Educational Experience (IPPEX)

NASA Astrophysics Data System (ADS)

Dominguez, A.; Rusaitis, L.; Zwicker, A.; Stotler, D. P.

2015-11-01

In the late 1990's PPPL's Science Education Department developed an innovative online site called the Interactive Plasma Physics Educational Experience (IPPEX). It featured (among other modules) two Java based applications which simulated tokamak physics: A steady state tokamak (SST) and a time dependent tokamak (TDT). The physics underlying the SST and the TDT are based on the ASPECT code which is a global power balance code developed to evaluate the performance of fusion reactor designs. We have relaunched the IPPEX site with updated modules and functionalities: The site itself is now dynamic on all platforms. The graphic design of the site has been modified to current standards. The virtual tokamak programming has been redone in Javascript, taking advantage of the speed and compactness of the code. The GUI of the tokamak has been completely redesigned, including more intuitive representations of changes in the plasma, e.g., particles moving along magnetic field lines. The use of GPU accelerated computation provides accurate and smooth visual representations of the plasma. We will present the current version of IPPEX as well near term plans of incorporating real time NSTX-U data into the simulation.

Science based integrated approach to advanced nuclear fuel development - integrated multi-scale multi-physics hierarchical modeling and simulation framework Part III: cladding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tome, Carlos N; Caro, J A; Lebensohn, R A

2010-01-01

Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Reactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems to develop predictive tools is critical. Not only are fabrication and performance models needed to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating themore » phase and microstructural behavior of the nuclear fuel system materials and matrices. In this paper we review the current status of the advanced modeling and simulation of nuclear reactor cladding, with emphasis on what is available and what is to be developed in each scale of the project, how we propose to pass information from one scale to the next, and what experimental information is required for benchmarking and advancing the modeling at each scale level.« less

143. ARAIII Control building (ARA607) Floor plan. Shows control room, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

143. ARA-III Control building (ARA-607) Floor plan. Shows control room, contaminated work area, counting and computer room, health physics room, instrument repair room, offices, and other rooms. Aerojet-general 880-area/GCRE-607-A-1. Date: February 1958. Ineel index code no. 063-0607-00-013-102546. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Gap Analysis of Material Properties Data for Ferritic/Martensitic HT-9 Steel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Neil R.; Serrano De Caro, Magdalena; Rodriguez, Edward A.

2012-08-28

The US Department of Energy (DOE), Office of Nuclear Energy (NE), is supporting the development of an ASME Code Case for adoption of 12Cr-1Mo-VW ferritic/martensitic (F/M) steel, commonly known as HT-9, primarily for use in elevated temperature design of liquid-metal fast reactors (LMFR) and components. In 2011, Los Alamos National Laboratory (LANL) nuclear engineering staff began assisting in the development of a small modular reactor (SMR) design concept, previously known as the Hyperion Module, now called the Gen4 Module. LANL staff immediately proposed HT-9 for the reactor vessel and components, as well as fuel clad and ducting, due to itsmore » superior thermal qualities. Although the ASME material Code Case, for adoption of HT-9 as an approved elevated temperature material for LMFR service, is the ultimate goal of this project, there are several key deliverables that must first be successfully accomplished. The most important key deliverable is the research, accumulation, and documentation of specific material parameters; physical, mechanical, and environmental, which becomes the basis for an ASME Code Case. Time-independent tensile and ductility data and time-dependent creep and creep-rupture behavior are some of the material properties required for a successful ASME Code case. Although this report provides a cursory review of the available data, a much more comprehensive study of open-source data would be necessary. This report serves three purposes: (a) provides a list of already existing material data information that could ultimately be made available to the ASME Code, (b) determines the HT-9 material properties data missing from available sources that would be required and (c) estimates the necessary material testing required to close the gap. Ultimately, the gap analysis demonstrates that certain material properties testing will be required to fulfill the necessary information package for an ASME Code Case.« less

Progress on China nuclear data processing code system

NASA Astrophysics Data System (ADS)

Liu, Ping; Wu, Xiaofei; Ge, Zhigang; Li, Songyang; Wu, Haicheng; Wen, Lili; Wang, Wenming; Zhang, Huanyu

2017-09-01

China is developing the nuclear data processing code Ruler, which can be used for producing multi-group cross sections and related quantities from evaluated nuclear data in the ENDF format [1]. The Ruler includes modules for reconstructing cross sections in all energy range, generating Doppler-broadened cross sections for given temperature, producing effective self-shielded cross sections in unresolved energy range, calculating scattering cross sections in thermal energy range, generating group cross sections and matrices, preparing WIMS-D format data files for the reactor physics code WIMS-D [2]. Programming language of the Ruler is Fortran-90. The Ruler is tested for 32-bit computers with Windows-XP and Linux operating systems. The verification of Ruler has been performed by comparison with calculation results obtained by the NJOY99 [3] processing code. The validation of Ruler has been performed by using WIMSD5B code.

Development of a Model and Computer Code to Describe Solar Grade Silicon Production Processes

NASA Technical Reports Server (NTRS)

Srivastava, R.; Gould, R. K.

1979-01-01

The program aims at developing mathematical models and computer codes based on these models, which allow prediction of the product distribution in chemical reactors for converting gaseous silicon compounds to condensed-phase silicon. The major interest is in collecting silicon as a liquid on the reactor walls and other collection surfaces. Two reactor systems are of major interest, a SiCl4/Na reactor in which Si(l) is collected on the flow tube reactor walls and a reactor in which Si(l) droplets formed by the SiCl4/Na reaction are collected by a jet impingement method. During this quarter the following tasks were accomplished: (1) particle deposition routines were added to the boundary layer code; and (2) Si droplet sizes in SiCl4/Na reactors at temperatures below the dew point of Si are being calculated.

RAZORBACK - A Research Reactor Transient Analysis Code Version 1.0 - Volume 3: Verification and Validation Report.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Talley, Darren G.

2017-04-01

This report describes the work and results of the verification and validation (V&V) of the version 1.0 release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, the equation of motion for fuel element thermal expansion, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This V&V effort was intended to confirm that the code showsmore » good agreement between simulation and actual ACRR operations.« less

Improvement of Modeling HTGR Neutron Physics by Uncertainty Analysis with the Use of Cross-Section Covariance Information

NASA Astrophysics Data System (ADS)

Boyarinov, V. F.; Grol, A. V.; Fomichenko, P. A.; Ternovykh, M. Yu

2017-01-01

This work is aimed at improvement of HTGR neutron physics design calculations by application of uncertainty analysis with the use of cross-section covariance information. Methodology and codes for preparation of multigroup libraries of covariance information for individual isotopes from the basic 44-group library of SCALE-6 code system were developed. A 69-group library of covariance information in a special format for main isotopes and elements typical for high temperature gas cooled reactors (HTGR) was generated. This library can be used for estimation of uncertainties, associated with nuclear data, in analysis of HTGR neutron physics with design codes. As an example, calculations of one-group cross-section uncertainties for fission and capture reactions for main isotopes of the MHTGR-350 benchmark, as well as uncertainties of the multiplication factor (k∞) for the MHTGR-350 fuel compact cell model and fuel block model were performed. These uncertainties were estimated by the developed technology with the use of WIMS-D code and modules of SCALE-6 code system, namely, by TSUNAMI, KENO-VI and SAMS. Eight most important reactions on isotopes for MHTGR-350 benchmark were identified, namely: 10B(capt), 238U(n,γ), ν5, 235U(n,γ), 238U(el), natC(el), 235U(fiss)-235U(n,γ), 235U(fiss).

Advances in Monte-Carlo code TRIPOLI-4®'s treatment of the electromagnetic cascade

NASA Astrophysics Data System (ADS)

Mancusi, Davide; Bonin, Alice; Hugot, François-Xavier; Malouch, Fadhel

2018-01-01

TRIPOLI-4® is a Monte-Carlo particle-transport code developed at CEA-Saclay (France) that is employed in the domains of nuclear-reactor physics, criticality-safety, shielding/radiation protection and nuclear instrumentation. The goal of this paper is to report on current developments, validation and verification made in TRIPOLI-4 in the electron/positron/photon sector. The new capabilities and improvements concern refinements to the electron transport algorithm, the introduction of a charge-deposition score, the new thick-target bremsstrahlung option, the upgrade of the bremsstrahlung model and the improvement of electron angular straggling at low energy. The importance of each of the developments above is illustrated by comparisons with calculations performed with other codes and with experimental data.

MC 2 -3: Multigroup Cross Section Generation Code for Fast Reactor Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, Changho; Yang, Won Sik

This paper presents the methods and performance of the MC2 -3 code, which is a multigroup cross-section generation code for fast reactor analysis, developed to improve the resonance self-shielding and spectrum calculation methods of MC2 -2 and to simplify the current multistep schemes generating region-dependent broad-group cross sections. Using the basic neutron data from ENDF/B data files, MC2 -3 solves the consistent P1 multigroup transport equation to determine the fundamental mode spectra for use in generating multigroup neutron cross sections. A homogeneous medium or a heterogeneous slab or cylindrical unit cell problem is solved in ultrafine (2082) or hyperfine (~400more » 000) group levels. In the resolved resonance range, pointwise cross sections are reconstructed with Doppler broadening at specified temperatures. The pointwise cross sections are directly used in the hyperfine group calculation, whereas for the ultrafine group calculation, self-shielded cross sections are prepared by numerical integration of the pointwise cross sections based upon the narrow resonance approximation. For both the hyperfine and ultrafine group calculations, unresolved resonances are self-shielded using the analytic resonance integral method. The ultrafine group calculation can also be performed for a two-dimensional whole-core problem to generate region-dependent broad-group cross sections. Verification tests have been performed using the benchmark problems for various fast critical experiments including Los Alamos National Laboratory critical assemblies; Zero-Power Reactor, Zero-Power Physics Reactor, and Bundesamt für Strahlenschutz experiments; Monju start-up core; and Advanced Burner Test Reactor. Verification and validation results with ENDF/B-VII.0 data indicated that eigenvalues from MC2 -3/DIF3D agreed well with Monte Carlo N-Particle5 MCNP5 or VIM Monte Carlo solutions within 200 pcm and regionwise one-group fluxes were in good agreement with Monte Carlo solutions.« less

3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Langenbuch, S.; Velkov, K.; Lizorkin, M.

1997-07-01

This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

SOURCELESS STARTUP. A MACHINE CODE FOR COMPUTING LOW-SOURCE REACTOR STARTUPS

DOE Office of Scientific and Technical Information (OSTI.GOV)

MacMillan, D.B.

1960-06-01

>A revision to the sourceless start-up code is presented. The code solves a system of differential equations encountered in computing the probability distribution of activity at an observed power level during reactor start-up from a very low source level. (J.R.D.)

Comparative study between single core model and detail core model of CFD modelling on reactor core cooling behaviour

NASA Astrophysics Data System (ADS)

Darmawan, R.

2018-01-01

Nuclear power industry is facing uncertainties since the occurrence of the unfortunate accident at Fukushima Daiichi Nuclear Power Plant. The issue of nuclear power plant safety becomes the major hindrance in the planning of nuclear power program for new build countries. Thus, the understanding of the behaviour of reactor system is very important to ensure the continuous development and improvement on reactor safety. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last four decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. Recently, 2D and 3D system level codes such as CFD are being explored. This paper discusses a comparative study on two different approaches of CFD modelling on reactor core cooling behaviour.

BISON and MARMOT Development for Modeling Fast Reactor Fuel Performance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gamble, Kyle Allan Lawrence; Williamson, Richard L.; Schwen, Daniel

2015-09-01

BISON and MARMOT are two codes under development at the Idaho National Laboratory for engineering scale and lower length scale fuel performance modeling. It is desired to add capabilities for fast reactor applications to these codes. The fast reactor fuel types under consideration are metal (U-Pu-Zr) and oxide (MOX). The cladding types of interest include 316SS, D9, and HT9. The purpose of this report is to outline the proposed plans for code development and provide an overview of the models added to the BISON and MARMOT codes for fast reactor fuel behavior. A brief overview of preliminary discussions on themore » formation of a bilateral agreement between the Idaho National Laboratory and the National Nuclear Laboratory in the United Kingdom is presented.« less

Physics-based multiscale coupling for full core nuclear reactor simulation

DOE PAGES

Gaston, Derek R.; Permann, Cody J.; Peterson, John W.; ...

2015-10-01

Numerical simulation of nuclear reactors is a key technology in the quest for improvements in efficiency, safety, and reliability of both existing and future reactor designs. Historically, simulation of an entire reactor was accomplished by linking together multiple existing codes that each simulated a subset of the relevant multiphysics phenomena. Recent advances in the MOOSE (Multiphysics Object Oriented Simulation Environment) framework have enabled a new approach: multiple domain-specific applications, all built on the same software framework, are efficiently linked to create a cohesive application. This is accomplished with a flexible coupling capability that allows for a variety of different datamore » exchanges to occur simultaneously on high performance parallel computational hardware. Examples based on the KAIST-3A benchmark core, as well as a simplified Westinghouse AP-1000 configuration, demonstrate the power of this new framework for tackling—in a coupled, multiscale manner—crucial reactor phenomena such as CRUD-induced power shift and fuel shuffle. 2014 The Authors. Published by Elsevier Ltd. This is an open access article under the CC BY-NC-SA license« less

Development of NSSS Thermal-Hydraulic Model for KNPEC-2 Simulator Using the Best-Estimate Code RETRAN-3D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Kyung-Doo; Jeong, Jae-Jun; Lee, Seung-Wook

The Nuclear Steam Supply System (NSSS) thermal-hydraulic model adopted in the Korea Nuclear Plant Education Center (KNPEC)-2 simulator was provided in the early 1980s. The reference plant for KNPEC-2 is the Yong Gwang Nuclear Unit 1, which is a Westinghouse-type 3-loop, 950 MW(electric) pressurized water reactor. Because of the limited computational capability at that time, it uses overly simplified physical models and assumptions for a real-time simulation of NSSS thermal-hydraulic transients. This may entail inaccurate results and thus, the possibility of so-called ''negative training,'' especially for complicated two-phase flows in the reactor coolant system. To resolve the problem, we developedmore » a realistic NSSS thermal-hydraulic program (named ARTS code) based on the best-estimate code RETRAN-3D. The systematic assessment of ARTS has been conducted by both a stand-alone test and an integrated test in the simulator environment. The non-integrated stand-alone test (NIST) results were reasonable in terms of accuracy, real-time simulation capability, and robustness. After successful completion of the NIST, ARTS was integrated with a 3-D reactor kinetics model and other system models. The site acceptance test (SAT) has been completed successively and confirmed to comply with the ANSI/ANS-3.5-1998 simulator software performance criteria. This paper presents our efforts for the ARTS development and some test results of the NIST and SAT.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vins, M.

This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe.more » Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)« less

Extensions of the MCNP5 and TRIPOLI4 Monte Carlo Codes for Transient Reactor Analysis

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard; Sjenitzer, Bart L.

2014-06-01

To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branchless collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3x3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3x3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail.

Coupling between a multi-physics workflow engine and an optimization framework

NASA Astrophysics Data System (ADS)

Di Gallo, L.; Reux, C.; Imbeaux, F.; Artaud, J.-F.; Owsiak, M.; Saoutic, B.; Aiello, G.; Bernardi, P.; Ciraolo, G.; Bucalossi, J.; Duchateau, J.-L.; Fausser, C.; Galassi, D.; Hertout, P.; Jaboulay, J.-C.; Li-Puma, A.; Zani, L.

2016-03-01

A generic coupling method between a multi-physics workflow engine and an optimization framework is presented in this paper. The coupling architecture has been developed in order to preserve the integrity of the two frameworks. The objective is to provide the possibility to replace a framework, a workflow or an optimizer by another one without changing the whole coupling procedure or modifying the main content in each framework. The coupling is achieved by using a socket-based communication library for exchanging data between the two frameworks. Among a number of algorithms provided by optimization frameworks, Genetic Algorithms (GAs) have demonstrated their efficiency on single and multiple criteria optimization. Additionally to their robustness, GAs can handle non-valid data which may appear during the optimization. Consequently GAs work on most general cases. A parallelized framework has been developed to reduce the time spent for optimizations and evaluation of large samples. A test has shown a good scaling efficiency of this parallelized framework. This coupling method has been applied to the case of SYCOMORE (SYstem COde for MOdeling tokamak REactor) which is a system code developed in form of a modular workflow for designing magnetic fusion reactors. The coupling of SYCOMORE with the optimization platform URANIE enables design optimization along various figures of merit and constraints.

Fuel Fraction Analysis of 500 MWth Gas Cooled Fast Reactor with Nitride (UN-PuN) Fuel without Refueling

NASA Astrophysics Data System (ADS)

Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

2017-01-01

Nuclear Power Plant (NPP) is one of candidates which can support electricity demand in the world. The Generation IV NPP has fourth main objective, i.e. sustainability, economics competitiveness, safety and reliability, and proliferation and physical protection. One of Gen-IV reactor type is Gas Cooled Fast Reactor (GFR). In this study, the analysis of fuel fraction in small GFR with nitride fuel has been done. The calculation was performed by SRAC code, both Pij and CITATION calculation. SRAC2002 system is a code system applicable to analyze the neutronics of variety reactor type. And for the data library used JENDL-3.2. The step of SRAC calculation is fuel pin calculated by Pij calculation until the data homogenized, after it homogenized we calculate core reactor. The variation of fuel fraction is 40% up to 65%. The optimum design of 500MWth GFR without refueling with 10 years burn up time reach when radius F1:F2:F3 = 50cm:30cm:30cm and height F1:F2:F3 = 50cm:40cm:30cm, variation percentage Plutonium in F1:F2:F3 = 7%:10%:13%. The optimum fuel fraction is 41% with addition 2% Plutonium weapon grade mix in the fuel. The excess reactivity value in this case 1.848% and the k-eff value is 1.01883. The high burn up reached when the fuel fraction is low. In this study 41% fuel fraction produce faster fissile fuel, so it has highest burn-up level than the other fuel fraction.

Computer optimization of reactor-thermoelectric space power systems

NASA Technical Reports Server (NTRS)

Maag, W. L.; Finnegan, P. M.; Fishbach, L. H.

1973-01-01

A computer simulation and optimization code that has been developed for nuclear space power systems is described. The results of using this code to analyze two reactor-thermoelectric systems are presented.

Optimization of coupled multiphysics methodology for safety analysis of pebble bed modular reactor

NASA Astrophysics Data System (ADS)

Mkhabela, Peter Tshepo

The research conducted within the framework of this PhD thesis is devoted to the high-fidelity multi-physics (based on neutronics/thermal-hydraulics coupling) analysis of Pebble Bed Modular Reactor (PBMR), which is a High Temperature Reactor (HTR). The Next Generation Nuclear Plant (NGNP) will be a HTR design. The core design and safety analysis methods are considerably less developed and mature for HTR analysis than those currently used for Light Water Reactors (LWRs). Compared to LWRs, the HTR transient analysis is more demanding since it requires proper treatment of both slower and much longer transients (of time scale in hours and days) and fast and short transients (of time scale in minutes and seconds). There is limited operation and experimental data available for HTRs for validation of coupled multi-physics methodologies. This PhD work developed and verified reliable high fidelity coupled multi-physics models subsequently implemented in robust, efficient, and accurate computational tools to analyse the neutronics and thermal-hydraulic behaviour for design optimization and safety evaluation of PBMR concept The study provided a contribution to a greater accuracy of neutronics calculations by including the feedback from thermal hydraulics driven temperature calculation and various multi-physics effects that can influence it. Consideration of the feedback due to the influence of leakage was taken into account by development and implementation of improved buckling feedback models. Modifications were made in the calculation procedure to ensure that the xenon depletion models were accurate for proper interpolation from cross section tables. To achieve this, the NEM/THERMIX coupled code system was developed to create the system that is efficient and stable over the duration of transient calculations that last over several tens of hours. Another achievement of the PhD thesis was development and demonstration of full-physics, three-dimensional safety analysis methodology for the PBMR to provide reference solutions. Investigation of different aspects of the coupled methodology and development of efficient kinetics treatment for the PBMR were carried out, which accounts for all feedback phenomena in an efficient manner. The OECD/NEA PBMR-400 coupled code benchmark was used as a test matrix for the proposed investigations. The integrated thermal-hydraulics and neutronics (multi-physics) methods were extended to enable modeling of a wider range of transients pertinent to the PBMR. First, the effect of the spatial mapping schemes (spatial coupling) was studied and quantified for different types of transients, which resulted in implementation of improved mapping methodology based on user defined criteria. The second aspect that was studied and optimized is the temporal coupling and meshing schemes between the neutronics and thermal-hydraulics time step selection algorithms. The coupled code convergence was achieved supplemented by application of methods to accelerate it. Finally, the modeling of all feedback phenomena in PBMRs was investigated and a novel treatment of cross-section dependencies was introduced for improving the representation of cross-section variations. The added benefit was that in the process of studying and improving the coupled multi-physics methodology more insight was gained into the physics and dynamics of PBMR, which will help also to optimize the PBMR design and improve its safety. One unique contribution of the PhD research is the investigation of the importance of the correct representation of the three-dimensional (3-D) effects in the PBMR analysis. The performed studies demonstrated that explicit 3-D modeling of control rod movement is superior and removes the errors associated with the grey curtain (2-D homogenized) approximation.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sen, Ramazan Sonat; Hummel, Andrew John; Hiruta, Hikaru

The deterministic full core simulators require homogenized group constants covering the operating and transient conditions over the entire lifetime. Traditionally, the homogenized group constants are generated using lattice physics code over an assembly or block in the case of prismatic high temperature reactors (HTR). For the case of strong absorbers that causes strong local depressions on the flux profile require special techniques during homogenization over a large volume. Fuel blocks with burnable poisons or control rod blocks are example of such cases. Over past several decades, there have been a tremendous number of studies performed for improving the accuracy ofmore » full-core calculations through the homogenization procedure. However, those studies were mostly performed for light water reactor (LWR) analyses, thus, may not be directly applicable to advanced thermal reactors such as HTRs. This report presents the application of SuPer-Homogenization correction method to a hypothetical HTR core.« less

H-division quarterly report, October--December 1977. [Lawrence Livermore Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1978-02-10

The Theoretical EOS Group develops theoretical techniques for describing material properties under extreme conditions and constructs equation-of-state (EOS) tables for specific applications. Work this quarter concentrated on a Li equation of state, equation of state for equilibrium plasma, improved ion corrections to the Thomas--Fermi--Kirzhnitz theory, and theoretical estimates of high-pressure melting in metals. The Experimental Physics Group investigates properties of materials at extreme conditions of pressure and temperature, and develops new experimental techniques. Effort this quarter concerned the following: parabolic projectile distortion in the two-state light-gas gun, construction of a ballistic range for long-rod penetrators, thermodynamics and sound velocities inmore » liquid metals, isobaric expansion measurements in Pt, and calculation of the velocity--mass profile of a jet produced by a shaped charge. Code development was concentrated on the PELE code, a multimaterial, multiphase, explicit finite-difference Eulerian code for pool suppression dynamics of a hypothetical loss-of-coolant accident in a nuclear reactor. Activities of the Fluid Dynamics Group were directed toward development of a code to compute the equations of state and transport properties of liquid metals (e.g. Li) and partially ionized dense plasmas, jet stability in the Li reactor system, and the study and problem application of fluid dynamic turbulence theory. 19 figures, 5 tables. (RWR)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gehin, Jess C; Godfrey, Andrew T; Evans, Thomas M

The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing a collection of methods and software products known as VERA, the Virtual Environment for Reactor Applications, including a core simulation capability called VERA-CS. A key milestone for this endeavor is to validate VERA against measurements from operating nuclear power reactors. The first step in validation against plant data is to determine the ability of VERA to accurately simulate the initial startup physics tests for Watts Bar Nuclear Power Station, Unit 1 (WBN1) cycle 1. VERA-CS calculations were performed with the Insilico code developed at ORNL using cross sectionmore » processing from the SCALE system and the transport capabilities within the Denovo transport code using the SPN method. The calculations were performed with ENDF/B-VII.0 cross sections in 252 groups (collapsed to 23 groups for the 3D transport solution). The key results of the comparison of calculations with measurements include initial criticality, control rod worth critical configurations, control rod worth, differential boron worth, and isothermal temperature reactivity coefficient (ITC). The VERA results for these parameters show good agreement with measurements, with the exception of the ITC, which requires additional investigation. Results are also compared to those obtained with Monte Carlo methods and a current industry core simulator.« less

Simulator platform for fast reactor operation and safety technology demonstration

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vilim, R. B.; Park, Y. S.; Grandy, C.

2012-07-30

A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe responsemore » to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.« less

Simulation of Nuclear Reactor Kinetics by the Monte Carlo Method

NASA Astrophysics Data System (ADS)

Gomin, E. A.; Davidenko, V. D.; Zinchenko, A. S.; Kharchenko, I. K.

2017-12-01

The KIR computer code intended for calculations of nuclear reactor kinetics using the Monte Carlo method is described. The algorithm implemented in the code is described in detail. Some results of test calculations are given.

ASME Code Efforts Supporting HTGRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

D.K. Morton

2010-09-01

In 1999, an international collaborative initiative for the development of advanced (Generation IV) reactors was started. The idea behind this effort was to bring nuclear energy closer to the needs of sustainability, to increase proliferation resistance, and to support concepts able to produce energy (both electricity and process heat) at competitive costs. The U.S. Department of Energy has supported this effort by pursuing the development of the Next Generation Nuclear Plant, a high temperature gas-cooled reactor. This support has included research and development of pertinent data, initial regulatory discussions, and engineering support of various codes and standards development. This reportmore » discusses the various applicable American Society of Mechanical Engineers (ASME) codes and standards that are being developed to support these high temperature gascooled reactors during construction and operation. ASME is aggressively pursuing these codes and standards to support an international effort to build the next generation of advanced reactors so that all can benefit.« less

ASME Code Efforts Supporting HTGRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

D.K. Morton

2011-09-01

In 1999, an international collaborative initiative for the development of advanced (Generation IV) reactors was started. The idea behind this effort was to bring nuclear energy closer to the needs of sustainability, to increase proliferation resistance, and to support concepts able to produce energy (both electricity and process heat) at competitive costs. The U.S. Department of Energy has supported this effort by pursuing the development of the Next Generation Nuclear Plant, a high temperature gas-cooled reactor. This support has included research and development of pertinent data, initial regulatory discussions, and engineering support of various codes and standards development. This reportmore » discusses the various applicable American Society of Mechanical Engineers (ASME) codes and standards that are being developed to support these high temperature gascooled reactors during construction and operation. ASME is aggressively pursuing these codes and standards to support an international effort to build the next generation of advanced reactors so that all can benefit.« less

ASME Code Efforts Supporting HTGRs

DOE Office of Scientific and Technical Information (OSTI.GOV)

D.K. Morton

2012-09-01

In 1999, an international collaborative initiative for the development of advanced (Generation IV) reactors was started. The idea behind this effort was to bring nuclear energy closer to the needs of sustainability, to increase proliferation resistance, and to support concepts able to produce energy (both electricity and process heat) at competitive costs. The U.S. Department of Energy has supported this effort by pursuing the development of the Next Generation Nuclear Plant, a high temperature gas-cooled reactor. This support has included research and development of pertinent data, initial regulatory discussions, and engineering support of various codes and standards development. This reportmore » discusses the various applicable American Society of Mechanical Engineers (ASME) codes and standards that are being developed to support these high temperature gascooled reactors during construction and operation. ASME is aggressively pursuing these codes and standards to support an international effort to build the next generation of advanced reactors so that all can benefit.« less

Verification of Three Dimensional Triangular Prismatic Discrete Ordinates Transport Code ENSEMBLE-TRIZ by Comparison with Monte Carlo Code GMVP

NASA Astrophysics Data System (ADS)

Homma, Yuto; Moriwaki, Hiroyuki; Ohki, Shigeo; Ikeda, Kazumi

2014-06-01

This paper deals with verification of three dimensional triangular prismatic discrete ordinates transport calculation code ENSEMBLE-TRIZ by comparison with multi-group Monte Carlo calculation code GMVP in a large fast breeder reactor. The reactor is a 750 MWe electric power sodium cooled reactor. Nuclear characteristics are calculated at beginning of cycle of an initial core and at beginning and end of cycle of equilibrium core. According to the calculations, the differences between the two methodologies are smaller than 0.0002 Δk in the multi-plication factor, relatively about 1% in the control rod reactivity, and 1% in the sodium void reactivity.

Thermal hydraulic-severe accident code interfaces for SCDAP/RELAP5/MOD3.2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Coryell, E.W.; Siefken, L.J.; Harvego, E.A.

1997-07-01

The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The code is the result of merging the RELAP5, SCDAP, and COUPLE codes. The RELAP5 portion of the code calculates the overall reactor coolant system, thermal-hydraulics, and associated reactor system responses. The SCDAP portion of the code describes the response of the core and associated vessel structures.more » The COUPLE portion of the code describes response of lower plenum structures and debris and the failure of the lower head. The code uses a modular approach with the overall structure, input/output processing, and data structures following the pattern established for RELAP5. The code uses a building block approach to allow the code user to easily represent a wide variety of systems and conditions through a powerful input processor. The user can represent a wide variety of experiments or reactor designs by selecting fuel rods and other assembly structures from a range of representative core component models, and arrange them in a variety of patterns within the thermalhydraulic network. The COUPLE portion of the code uses two-dimensional representations of the lower plenum structures and debris beds. The flow of information between the different portions of the code occurs at each system level time step advancement. The RELAP5 portion of the code describes the fluid transport around the system. These fluid conditions are used as thermal and mass transport boundary conditions for the SCDAP and COUPLE structures and debris beds.« less

Methodology comparison for gamma-heating calculations in material-testing reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lemaire, M.; Vaglio-Gaudard, C.; Lyoussi, A.

2015-07-01

The Jules Horowitz Reactor (JHR) is a Material-Testing Reactor (MTR) under construction in the south of France at CEA Cadarache (French Alternative Energies and Atomic Energy Commission). It will typically host about 20 simultaneous irradiation experiments in the core and in the beryllium reflector. These experiments will help us better understand the complex phenomena occurring during the accelerated ageing of materials and the irradiation of nuclear fuels. Gamma heating, i.e. photon energy deposition, is mainly responsible for temperature rise in non-fuelled zones of nuclear reactors, including JHR internal structures and irradiation devices. As temperature is a key parameter for physicalmore » models describing the behavior of material, accurate control of temperature, and hence gamma heating, is required in irradiation devices and samples in order to perform an advanced suitable analysis of future experimental results. From a broader point of view, JHR global attractiveness as a MTR depends on its ability to monitor experimental parameters with high accuracy, including gamma heating. Strict control of temperature levels is also necessary in terms of safety. As JHR structures are warmed up by gamma heating, they must be appropriately cooled down to prevent creep deformation or melting. Cooling-power sizing is based on calculated levels of gamma heating in the JHR. Due to these safety concerns, accurate calculation of gamma heating with well-controlled bias and associated uncertainty as low as possible is all the more important. There are two main kinds of calculation bias: bias coming from nuclear data on the one hand and bias coming from physical approximations assumed by computer codes and by general calculation route on the other hand. The former must be determined by comparison between calculation and experimental data; the latter by calculation comparisons between codes and between methodologies. In this presentation, we focus on this latter kind of bias. Nuclear heating is represented by the physical quantity called absorbed dose (energy deposition induced by particle-matter interactions, divided by mass). Its calculation with Monte Carlo codes is possible but computationally expensive as it requires transport simulation of charged particles, along with neutrons and photons. For that reason, the calculation of another physical quantity, called KERMA, is often preferred, as KERMA calculation with Monte Carlo codes only requires transport of neutral particles. However, KERMA is only an estimator of the absorbed dose and many conditions must be fulfilled for KERMA to be equal to absorbed dose, including so-called condition of electronic equilibrium. Also, Monte Carlo computations of absorbed dose still present some physical approximations, even though there is only a limited number of them. Some of these approximations are linked to the way how Monte Carlo codes apprehend the transport simulation of charged particles and the productive and destructive interactions between photons, electrons and positrons. There exists a huge variety of electromagnetic shower models which tackle this topic. Differences in the implementation of these models can lead to discrepancies in calculated values of absorbed dose between different Monte Carlo codes. The magnitude of order of such potential discrepancies should be quantified for JHR gamma-heating calculations. We consequently present a two-pronged plan. In a first phase, we intend to perform compared absorbed dose / KERMA Monte Carlo calculations in the JHR. This way, we will study the presence or absence of electronic equilibrium in the different JHR structures and experimental devices and we will give recommendations for the choice of KERMA or absorbed dose when calculating gamma heating in the JHR. In a second phase, we intend to perform compared TRIPOLI4 / MCNP absorbed dose calculations in a simplified JHR-representative geometry. For this comparison, we will use the same nuclear data library for both codes (the European library JEFF3.1.1 and photon library EPDL97) so as to isolate the effects from electromagnetic shower models on absorbed dose calculation. This way, we hope to get insightful feedback on these models and their implementation in Monte Carlo codes. (authors)« less

Adaptive Nodal Transport Methods for Reactor Transient Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Thomas Downar; E. Lewis

2005-08-31

Develop methods for adaptively treating the angular, spatial, and time dependence of the neutron flux in reactor transient analysis. These methods were demonstrated in the DOE transport nodal code VARIANT and the US NRC spatial kinetics code, PARCS.

Coupled neutronics and thermal-hydraulics numerical simulations of a Molten Fast Salt Reactor (MFSR)

NASA Astrophysics Data System (ADS)

Laureau, A.; Rubiolo, P. R.; Heuer, D.; Merle-Lucotte, E.; Brovchenko, M.

2014-06-01

Coupled neutronics and thermalhydraulic numerical analyses of a molten salt fast reactor are presented. These preliminary numerical simulations are carried-out using the Monte Carlo code MCNP and the Computation Fluid Dynamic code OpenFOAM. The main objectives of this analysis performed at steady-reactor conditions are to confirm the acceptability of the current neutronic and thermalhydraulic designs of the reactor, to study the effects of the reactor operating conditions on some of the key MSFR design parameters such as the temperature peaking factor. The effects of the precursor's motion on the reactor safety parameters such as the effective fraction of delayed neutrons have been evaluated.

VERA and VERA-EDU 3.5 Release Notes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sieger, Matt; Salko, Robert K.; Kochunas, Brendan M.

The Virtual Environment for Reactor Applications components included in this distribution include selected computational tools and supporting infrastructure that solve neutronics, thermal-hydraulics, fuel performance, and coupled neutronics-thermal hydraulics problems. The infrastructure components provide a simplified common user input capability and provide for the physics integration with data transfer and coupled-physics iterative solution algorithms. Neutronics analysis can be performed for 2D lattices, 2D core and 3D core problems for pressurized water reactor geometries that can be used to calculate criticality and fission rate distributions by pin for input fuel compositions. MPACT uses the Method of Characteristics transport approach for 2D problems.more » For 3D problems, MPACT uses the 2D/1D method which uses 2D MOC in a radial plane and diffusion or SPn in the axial direction. MPACT includes integrated cross section capabilities that provide problem-specific cross sections generated using the subgroup methodology. The code can be executed both 2D and 3D problems in parallel to reduce overall run time. A thermal-hydraulics capability is provided with CTF (an updated version of COBRA-TF) that allows thermal-hydraulics analyses for single and multiple assemblies using the simplified VERA common input. This distribution also includes coupled neutronics/thermal-hydraulics capabilities to allow calculations using MPACT coupled with CTF. The VERA fuel rod performance component BISON calculates, on a 2D or 3D basis, fuel rod temperature, fuel rod internal pressure, free gas volume, clad integrity and fuel rod waterside diameter. These capabilities allow simulation of power cycling, fuel conditioning and deconditioning, high burnup performance, power uprate scoping studies, and accident performance. Input/Output capabilities include the VERA Common Input (VERAIn) script which converts the ASCII common input file to the intermediate XML used to drive all of the physics codes in the VERA Core Simulator (VERA-CS). VERA component codes either input the VERA XML format directly, or provide a preprocessor which can convert the XML into native input. VERAView is an interactive graphical interface for the visualization and engineering analyses of output data from VERA. The python-based software is easy to install and intuitive to use, and provides instantaneous 2D and 3D images, 1D plots, and alpha-numeric data from VERA multi-physics simulations. Testing within CASL has focused primarily on Westinghouse four-loop reactor geometries and conditions with example problems included in the distribution.« less

Validation of DRAGON4/DONJON4 simulation methodology for a typical MNSR by calculating reactivity feedback coefficient and neutron flux

NASA Astrophysics Data System (ADS)

Al Zain, Jamal; El Hajjaji, O.; El Bardouni, T.; Boukhal, H.; Jaï, Otman

2018-06-01

The MNSR is a pool type research reactor, which is difficult to model because of the importance of neutron leakage. The aim of this study is to evaluate a 2-D transport model for the reactor compatible with the latest release of the DRAGON code and 3-D diffusion of the DONJON code. DRAGON code is then used to generate the group macroscopic cross sections needed for full core diffusion calculations. The diffusion DONJON code, is then used to compute the effective multiplication factor (keff), the feedback reactivity coefficients and neutron flux which account for variation in fuel and moderator temperatures as well as the void coefficient have been calculated using the DRAGON and DONJON codes for the MNSR research reactor. The cross sections of all the reactor components at different temperatures were generated using the DRAGON code. These group constants were used then in the DONJON code to calculate the multiplication factor and the neutron spectrum at different water and fuel temperatures using 69 energy groups. Only one parameter was changed where all other parameters were kept constant. Finally, Good agreements between the calculated and measured have been obtained for every of the feedback reactivity coefficients and neutron flux.

Use of the ETA-1 reactor for the validation of the multi-group APOLLO2-MORET 5 code and the Monte Carlo continuous energy MORET 5 code

NASA Astrophysics Data System (ADS)

Leclaire, N.; Cochet, B.; Le Dauphin, F. X.; Haeck, W.; Jacquet, O.

2014-06-01

The present paper aims at providing experimental validation for the use of the MORET 5 code for advanced concepts of reactor involving thorium and heavy water. It therefore constitutes an opportunity to test and improve the thermal-scattering data of heavy water and also to test the recent implementation of probability tables in the MORET 5 code.

Prediction of the Reactor Antineutrino Flux for the Double Chooz Experiment

NASA Astrophysics Data System (ADS)

Jones, Chirstopher LaDon

This thesis benchmarks the deterministic lattice code, DRAGON, against data, and then applies this code to make a prediction for the antineutrino flux from the Chooz Bl and B2 reactors. Data from the destructive assay of rods from the Takahama-3 reactor and from the SONGS antineutrino detector are used for comparisons. The resulting prediction from the tuned DRAGON code is then compared to the first antineutrino event spectra from Double Chooz. Use of this simulation in nuclear nonproliferation studies is discussed. (Copies available exclusively from MIT Libraries, libraries.mit.edu/docs - docs@mit.edu)

One-dimensional MHD simulations of MTF systems with compact toroid targets and spherical liners

NASA Astrophysics Data System (ADS)

Khalzov, Ivan; Zindler, Ryan; Barsky, Sandra; Delage, Michael; Laberge, Michel

2017-10-01

One-dimensional (1D) MHD code is developed in General Fusion (GF) for coupled plasma-liner simulations in magnetized target fusion (MTF) systems. The main goal of these simulations is to search for optimal parameters of MTF reactor, in which spherical liquid metal liner compresses compact toroid plasma. The code uses Lagrangian description for both liner and plasma. The liner is represented as a set of spherical shells with fixed masses while plasma is discretized as a set of nested tori with circular cross sections and fixed number of particles between them. All physical fields are 1D functions of either spherical (liner) or small toroidal (plasma) radius. Motion of liner and plasma shells is calculated self-consistently based on applied forces and equations of state. Magnetic field is determined by 1D profiles of poloidal and toroidal fluxes - they are advected with shells and diffuse according to local resistivity, this also accounts for flux leakage into the liner. Different plasma transport models are implemented, this allows for comparison with ongoing GF experiments. Fusion power calculation is included into the code. We performed a series of parameter scans in order to establish the underlying dependencies of the MTF system and find the optimal reactor design point.

Overview and Current Status of Analyses of Potential LEU Design Concepts for TREAT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Connaway, H. M.; Kontogeorgakos, D. C.; Papadias, D. D.

2015-10-01

Neutronic and thermal-hydraulic analyses have been performed to evaluate the performance of different low-enriched uranium (LEU) fuel design concepts for the conversion of the Transient Reactor Test Facility (TREAT) from its current high-enriched uranium (HEU) fuel. TREAT is an experimental reactor developed to generate high neutron flux transients for the testing of nuclear fuels. The goal of this work was to identify an LEU design which can maintain the performance of the existing HEU core while continuing to operate safely. A wide variety of design options were considered, with a focus on minimizing peak fuel temperatures and optimizing the powermore » coupling between the TREAT core and test samples. Designs were also evaluated to ensure that they provide sufficient reactivity and shutdown margin for each control rod bank. Analyses were performed using the core loading and experiment configuration of historic M8 Power Calibration experiments (M8CAL). The Monte Carlo code MCNP was utilized for steady-state analyses, and transient calculations were performed with the point kinetics code TREKIN. Thermal analyses were performed with the COMSOL multi-physics code. Using the results of this study, a new LEU Baseline design concept is being established, which will be evaluated in detail in a future report.« less

Verification of a neutronic code for transient analysis in reactors with Hex-z geometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gonzalez-Pintor, S.; Verdu, G.; Ginestar, D.

Due to the geometry of the fuel bundles, to simulate reactors such as VVER reactors it is necessary to develop methods that can deal with hexagonal prisms as basic elements of the spatial discretization. The main features of a code based on a high order finite element method for the spatial discretization of the neutron diffusion equation and an implicit difference method for the time discretization of this equation are presented and the performance of the code is tested solving the first exercise of the AER transient benchmark. The obtained results are compared with the reference results of the benchmarkmore » and with the results provided by PARCS code. (authors)« less

Multidimensional Fuel Performance Code: BISON

DOE Office of Scientific and Technical Information (OSTI.GOV)

BISON is a finite element based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO fuel particles, and metallic rod and plate fuel (Refs. [a, b, c]). It solves the fully-coupled equations of thermomechanics and species diffusion and includes important fuel physics such as fission gas release and material property degradation with burnup. BISON is based on the MOOSE framework (Ref. [d]) and can therefore efficiently solve problems on 1-, 2- or 3-D meshes using standard workstations or large high performance computers. BISON is also coupled to a MOOSE-based mesoscale phasemore » field material property simulation capability (Refs. [e, f]). As described here, BISON includes the code library named FOX, which was developed concurrent with BISON. FOX contains material and behavioral models that are specific to oxide fuels.« less

Decay heat of sodium fast reactor: Comparison of experimental measurements on the PHENIX reactor with calculations performed with the French DARWIN package

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benoit, J. C.; Bourdot, P.; Eschbach, R.

2012-07-01

A Decay Heat (DH) experiment on the whole core of the French Sodium-Cooled Fast Reactor PHENIX has been conducted in May 2008. The measurements began an hour and a half after the shutdown of the reactor and lasted twelve days. It is one of the experiments used for the experimental validation of the depletion code DARWIN thereby confirming the excellent performance of the aforementioned code. Discrepancies between measured and calculated decay heat do not exceed 8%. (authors)

Assessment and Application of the ROSE Code for Reactor Outage Thermal-Hydraulic and Safety Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liang, Thomas K.S.; Ko, F.-K.; Dai, L.-C

The currently available tools, such as RELAP5, RETRAN, and others, cannot easily and correctly perform the task of analyzing the system behavior during plant outages. Therefore, a medium-sized program aiming at reactor outage simulation and evaluation, such as midloop operation (MLO) with loss of residual heat removal (RHR), has been developed. Important thermal-hydraulic processes involved during MLO with loss of RHR can be properly simulated by the newly developed reactor outage simulation and evaluation (ROSE) code. The two-region approach with a modified two-fluid model has been adopted to be the theoretical basis of the ROSE code.To verify the analytical modelmore » in the first step, posttest calculations against the integral midloop experiments with loss of RHR have been performed. The excellent simulation capacity of the ROSE code against the Institute of Nuclear Energy Research Integral System Test Facility test data is demonstrated. To further mature the ROSE code in simulating a full-sized pressurized water reactor, assessment against the WGOTHIC code and the Maanshan momentary-loss-of-RHR event has been undertaken. The successfully assessed ROSE code is then applied to evaluate the abnormal operation procedure (AOP) with loss of RHR during MLO (AOP 537.4) for the Maanshan plant. The ROSE code also has been successfully transplanted into the Maanshan training simulator to support operator training. How the simulator was upgraded by the ROSE code for MLO will be presented in the future.« less

Multi-phase model development to assess RCIC system capabilities under severe accident conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kirkland, Karen Vierow; Ross, Kyle; Beeny, Bradley

The Reactor Core Isolation Cooling (RCIC) System is a safety-related system that provides makeup water for core cooling of some Boiling Water Reactors (BWRs) with a Mark I containment. The RCIC System consists of a steam-driven Terry turbine that powers a centrifugal, multi-stage pump for providing water to the reactor pressure vessel. The Fukushima Dai-ichi accidents demonstrated that the RCIC System can play an important role under accident conditions in removing core decay heat. The unexpectedly sustained, good performance of the RCIC System in the Fukushima reactor demonstrates, firstly, that its capabilities are not well understood, and secondly, that themore » system has high potential for extended core cooling in accident scenarios. Better understanding and analysis tools would allow for more options to cope with a severe accident situation and to reduce the consequences. The objectives of this project were to develop physics-based models of the RCIC System, incorporate them into a multi-phase code and validate the models. This Final Technical Report details the progress throughout the project duration and the accomplishments.« less

SCALE Code System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rearden, Bradley T.; Jessee, Matthew Anderson

The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministicmore » and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.« less

SCALE Code System 6.2.1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rearden, Bradley T.; Jessee, Matthew Anderson

The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministicmore » and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.« less

Development of fission-products transport model in severe-accident scenarios for Scdap/Relap5

NASA Astrophysics Data System (ADS)

Honaiser, Eduardo Henrique Rangel

The understanding and estimation of the release of fission products during a severe accident became one of the priorities of the nuclear community after 1980, with the events of the Three-mile Island unit 2 (TMI-2), in 1979, and Chernobyl accidents, in 1986. Since this time, theoretical developments and experiments have shown that the primary circuit systems of light water reactors (LWR) have the potential to attenuate the release of fission products, a fact that had been neglected before. An advanced tool, compatible with nuclear thermal-hydraulics integral codes, is developed to predict the retention and physical evolution of the fission products in the primary circuit of LWRs, without considering the chemistry effects. The tool embodies the state-of-the-art models for the involved phenomena as well as develops new models. The capabilities acquired after the implementation of this tool in the Scdap/Relap5 code can be used to increase the accuracy of probability safety assessment (PSA) level 2, enhance the reactor accident management procedures and design new emergency safety features.

On use of ZPR research reactors and associated instrumentation and measurement methods for reactor physics studies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chauvin, J.P.; Blaise, P.; Lyoussi, A.

2015-07-01

The French atomic and alternative energies -CEA- is strongly involved in research and development programs concerning the use of nuclear energy as a clean and reliable source of energy and consequently is working on the present and future generation of reactors on various topics such as ageing plant management, optimization of the plutonium stockpile, waste management and innovative systems exploration. Core physics studies are an essential part of this comprehensive R and D effort. In particular, the Zero Power Reactor (ZPR) of CEA: EOLE, MINERVE and MASURCA play an important role in the validation of neutron (as well photon) physicsmore » calculation tools (codes and nuclear data). The experimental programs defined in the CEA's ZPR facilities aim at improving the calculation routes by reducing the uncertainties of the experimental databases. They also provide accurate data on innovative systems in terms of new materials (moderating and decoupling materials) and new concepts (ADS, ABWR, new MTR (e.g. JHR), GENIV) involving new fuels, absorbers and coolant materials. Conducting such interesting experimental R and D programs is based on determining and measuring main parameters of phenomena of interest to qualify calculation tools and nuclear data 'libraries'. Determining these parameters relies on the use of numerous and different experimental techniques using specific and appropriate instrumentation and detection tools. Main ZPR experimental programs at CEA, their objectives and challenges will be presented and discussed. Future development and perspectives regarding ZPR reactors and associated programs will be also presented. (authors)« less

TRANSURANUS: a fuel rod analysis code ready for use

NASA Astrophysics Data System (ADS)

Lassmann, K.

1992-06-01

TRANSURANUS is a computer program for the thermal and mechanical analysis of fuel rods in nuclear reactors and was developed at the European Institute for Transuranium Elements (TUI). The TRANSURANUS code consists of a clearly defined mechanical-mathematical framework into which physical models can easily be incorporated. Besides its flexibility for different fuel rod designs the TRANSURANUS code can deal with very different situations, as given for instance in an experiment, under normal, off-normal and accident conditions. The time scale of the problems to be treated may range from milliseconds to years. The code has a comprehensive material data bank for oxide, mixed oxide, carbide and nitride fuels, Zircaloy and steel claddings and different coolants. During its development great effort was spent on obtaining an extremely flexible tool which is easy to handle, exhibiting very fast running times. The total development effort is approximately 40 man-years. In recent years the interest to use this code grew and the code is in use in several organisations, both research and private industry. The code is now available to all interested parties. The paper outlines the main features and capabilities of the TRANSURANUS code, its validation and treats also some practical aspects.

Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liao, J.; Kucukboyaci, V. N.; Nguyen, L.

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using themore » WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)« less

An investigation of tritium transfer in reactor loops

NASA Astrophysics Data System (ADS)

Ilyasova, O. H.; Mosunova, N. A.

2017-09-01

The work is devoted to the important task of the numerical simulation and analysis of the tritium behaviour in the reactor loops. The simulation was carried out by HYDRA-IBRAE/LM code, which is being developed in Nuclear safety institute of the Russian Academy of Sciences. The code is intended for modeling of the liquid metal flow (sodium, lead and lead-bismuth) on the base of non-homogeneous and non-equilibrium two-fluid model. In order to simulate tritium transfer in the code, the special module has been developed. Module includes the models describing the main phenomena of tritium behaviour in reactor loops: transfer, permeation, leakage, etc. Because of shortage of the experimental data, a lot of analytical tests and comparative calculations were considered. Some of them are presented in this work. The comparison of estimation results and experimental and analytical data demonstrate not only qualitative but also good quantitative agreement. It is possible to confirm that HYDRA-IBRAE/LM code allows modeling tritium transfer in reactor loops.

Safety Criticality Standards Using the French CRISTAL Code Package: Application to the AREVA NP UO{sub 2} Fuel Fabrication Plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Doucet, M.; Durant Terrasson, L.; Mouton, J.

2006-07-01

Criticality safety evaluations implement requirements to proof of sufficient sub critical margins outside of the reactor environment for example in fuel fabrication plants. Basic criticality data (i.e., criticality standards) are used in the determination of sub critical margins for all processes involving plutonium or enriched uranium. There are several criticality international standards, e.g., ARH-600, which is one the US nuclear industry relies on. The French Nuclear Safety Authority (DGSNR and its advising body IRSN) has requested AREVA NP to review the criticality standards used for the evaluation of its Low Enriched Uranium fuel fabrication plants with CRISTAL V0, the recentlymore » updated French criticality evaluation package. Criticality safety is a concern for every phase of the fabrication process including UF{sub 6} cylinder storage, UF{sub 6}-UO{sub 2} conversion, powder storage, pelletizing, rod loading, assembly fabrication, and assembly transportation. Until 2003, the accepted criticality standards were based on the French CEA work performed in the late seventies with the APOLLO1 cell/assembly computer code. APOLLO1 is a spectral code, used for evaluating the basic characteristics of fuel assemblies for reactor physics applications, which has been enhanced to perform criticality safety calculations. Throughout the years, CRISTAL, starting with APOLLO1 and MORET 3 (a 3D Monte Carlo code), has been improved to account for the growth of its qualification database and for increasing user requirements. Today, CRISTAL V0 is an up-to-date computational tool incorporating a modern basic microscopic cross section set based on JEF2.2 and the comprehensive APOLLO2 and MORET 4 codes. APOLLO2 is well suited for criticality standards calculations as it includes a sophisticated self shielding approach, a P{sub ij} flux determination, and a 1D transport (S{sub n}) process. CRISTAL V0 is the result of more than five years of development work focusing on theoretical approaches and the implementation of user-friendly graphical interfaces. Due to its comprehensive physical simulation and thanks to its broad qualification database with more than a thousand benchmark/calculation comparisons, CRISTAL V0 provides outstanding and reliable accuracy for criticality evaluations for configurations covering the entire fuel cycle (i.e. from enrichment, pellet/assembly fabrication, transportation, to fuel reprocessing). After a brief description of the calculation scheme and the physics algorithms used in this code package, results for the various fissile media encountered in a UO{sub 2} fuel fabrication plant will be detailed and discussed. (authors)« less

Development of a three-dimensional transient code for reactivity-initiated events of BWRs (boiling water reactors) - Models and code verifications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Uematsu, Hitoshi; Yamamoto, Toru; Izutsu, Sadayuki

1990-06-01

A reactivity-initiated event is a design-basis accident for the safety analysis of boiling water reactors. It is defined as a rapid transient of reactor power caused by a reactivity insertion of over $1.0 due to a postulated drop or abnormal withdrawal of the control rod from the core. Strong space-dependent feedback effects are associated with the local power increase due to control rod movement. A realistic treatment of the core status in a transient by a code with a detailed core model is recommended in evaluating this event. A three-dimensional transient code, ARIES, has been developed to meet this need.more » The code simulates the event with three-dimensional neutronics, coupled with multichannel thermal hydraulics, based on a nonequilibrium separated flow model. The experimental data obtained in reactivity accident tests performed with the SPERT III-E core are used to verify the entire code, including thermal-hydraulic models.« less

Application of CFX-10 to the Investigation of RPV Coolant Mixing in VVER Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moretti, Fabio; Melideo, Daniele; Terzuoli, Fulvio

2006-07-01

Coolant mixing phenomena occurring in the pressure vessel of a nuclear reactor constitute one of the main objectives of investigation by researchers concerned with nuclear reactor safety. For instance, mixing plays a relevant role in reactivity-induced accidents initiated by de-boration or boron dilution events, followed by transport of a de-borated slug into the vessel of a pressurized water reactor. Another example is constituted by temperature mixing, which may sensitively affect the consequences of a pressurized thermal shock scenario. Predictive analysis of mixing phenomena is strongly improved by the availability of computational tools able to cope with the inherent three-dimensionality ofmore » such problem, like system codes with three-dimensional capabilities, and Computational Fluid Dynamics (CFD) codes. The present paper deals with numerical analyses of coolant mixing in the reactor pressure vessel of a VVER-1000 reactor, performed by the ANSYS CFX-10 CFD code. In particular, the 'swirl' effect that has been observed to take place in the downcomer of such kind of reactor has been addressed, with the aim of assessing the capability of the codes to predict that effect, and to understand the reasons for its occurrence. Results have been compared against experimental data from V1000CT-2 Benchmark. Moreover, a boron mixing problem has been investigated, in the hypothesis that a de-borated slug, transported by natural circulation, enters the vessel. Sensitivity analyses have been conducted on some geometrical features, model parameters and boundary conditions. (authors)« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salko, Robert K; Sung, Yixing; Kucukboyaci, Vefa

The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time stepmore » of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.« less

Modeling and simulation of CANDU reactor and its regulating system

NASA Astrophysics Data System (ADS)

Javidnia, Hooman

Analytical computer codes are indispensable tools in design, optimization, and control of nuclear power plants. Numerous codes have been developed to perform different types of analyses related to the nuclear power plants. A large number of these codes are designed to perform safety analyses. In the context of safety analyses, the control system is often neglected. Although there are good reasons for such a decision, that does not mean that the study of control systems in the nuclear power plants should be neglected altogether. In this thesis, a proof of concept code is developed as a tool that can be used in the design. optimization. and operation stages of the control system. The main objective in the design of this computer code is providing a tool that is easy to use by its target audience and is capable of producing high fidelity results that can be trusted to design the control system and optimize its performance. Since the overall plant control system covers a very wide range of processes, in this thesis the focus has been on one particular module of the the overall plant control system, namely, the reactor regulating system. The center of the reactor regulating system is the CANDU reactor. A nodal model for the reactor is used to represent the spatial neutronic kinetics of the core. The nodal model produces better results compared to the point kinetics model which is often used in the design and analysis of control system for nuclear reactors. The model can capture the spatial effects to some extent. although it is not as detailed as the finite difference methods. The criteria for choosing a nodal model of the core are: (1) the model should provide more detail than point kinetics and capture spatial effects, (2) it should not be too complex or overly detailed to slow down the simulation and provide details that are extraneous or unnecessary for a control engineer. Other than the reactor itself, there are auxiliary models that describe dynamics of different phenomena related to the transfer of the energy from the core. The main function of the reactor regulating system is to control the power of the reactor. This is achieved by using a set of detectors. reactivity devices. and digital control algorithms. Three main reactivity devices that are activated during short-term or intermediate-term transients are modeled in this thesis. The main elements of the digital control system are implemented in accordance to the program specifications for the actual control system in CANDU reactors. The simulation results are validated against requirements of the reactor regulating system. actual plant data. and pre-validated data from other computer codes. The validation process shows that the simulation results can be trusted in making engineering decisions regarding the reactor regulating system and prediction of the system performance in response to upset conditions or disturbances. KEYWORDS: CANDU reactors. reactor regulating system. nodal model. spatial kinetics. reactivity devices. simulation.

MATCHED-INDEX-OF-REFRACTION FLOW FACILITY FOR FUNDAMENTAL AND APPLIED RESEARCH

DOE Office of Scientific and Technical Information (OSTI.GOV)

Piyush Sabharwall; Carl Stoots; Donald M. McEligot

2014-11-01

Significant challenges face reactor designers with regard to thermal hydraulic design and associated modeling for advanced reactor concepts. Computational thermal hydraulic codes solve only a piece of the core. There is a need for a whole core dynamics system code with local resolution to investigate and understand flow behavior with all the relevant physics and thermo-mechanics. The matched index of refraction (MIR) flow facility at Idaho National Laboratory (INL) has a unique capability to contribute to the development of validated computational fluid dynamics (CFD) codes through the use of state-of-the-art optical measurement techniques, such as Laser Doppler Velocimetry (LDV) andmore » Particle Image Velocimetry (PIV). PIV is a non-intrusive velocity measurement technique that tracks flow by imaging the movement of small tracer particles within a fluid. At the heart of a PIV calculation is the cross correlation algorithm, which is used to estimate the displacement of particles in some small part of the image over the time span between two images. Generally, the displacement is indicated by the location of the largest peak. To quantify these measurements accurately, sophisticated processing algorithms correlate the locations of particles within the image to estimate the velocity (Ref. 1). Prior to use with reactor deign, the CFD codes have to be experimentally validated, which requires rigorous experimental measurements to produce high quality, multi-dimensional flow field data with error quantification methodologies. Computational thermal hydraulic codes solve only a piece of the core. There is a need for a whole core dynamics system code with local resolution to investigate and understand flow behavior with all the relevant physics and thermo-mechanics. Computational techniques with supporting test data may be needed to address the heat transfer from the fuel to the coolant during the transition from turbulent to laminar flow, including the possibility of an early laminarization of the flow (Refs. 2 and 3) (laminarization is caused when the coolant velocity is theoretically in the turbulent regime, but the heat transfer properties are indicative of the coolant velocity being in the laminar regime). Such studies are complicated enough that computational fluid dynamics (CFD) models may not converge to the same conclusion. Thus, experimentally scaled thermal hydraulic data with uncertainties should be developed to support modeling and simulation for verification and validation activities. The fluid/solid index of refraction matching technique allows optical access in and around geometries that would otherwise be impossible while the large test section of the INL system provides better spatial and temporal resolution than comparable facilities. Benchmark data for assessing computational fluid dynamics can be acquired for external flows, internal flows, and coupled internal/external flows for better understanding of physical phenomena of interest. The core objective of this study is to describe MIR and its capabilities, and mention current development areas for uncertainty quantification, mainly the uncertainty surface method and cross-correlation method. Using these methods, it is anticipated to establish a suitable approach to quantify PIV uncertainty for experiments performed in the MIR.« less

Corrigendum to “Accelerated materials evaluation for nuclear applications” [J. Nucl. Mater. 488 (2017) 46–62

DOE PAGES

Griffiths, Malcolm; Walters, L.; Greenwood, L. R.; ...

2017-09-21

The original article addresses the opportunities and complexities of using materials test reactors with high neutron fluxes to perform accelerated studies of material aging in power reactors operating at lower neutron fluxes and with different neutron flux spectra. Radiation damage and gas production in different reactors have been compared using the code, SPECTER. This code provides a common standard from which to compare neutron damage data generated by different research groups using a variety of reactors. This Corrigendum identifies a few typographical errors. Tables 2 and 3 are included in revised form.

151. ARAIII Reactor building (ARA608) Details of reactor pit and ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

151. ARA-III Reactor building (ARA-608) Details of reactor pit and instrument plan. Aerojet-general 880-area/GCRE-608-T-19. Date: November 1958. Ineel index code no. 063-0608-25-013-102678. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user`s manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.

This document is the User`s Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code`s capabilities and limitations; Chapter 2 describes the code`s structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARCmore » and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs.« less

Double differential light charged particle emission cross sections for some structural fusion materials

NASA Astrophysics Data System (ADS)

Sarpün, Ismail Hakki; n, Abdullah Aydı; Tel, Eyyup

2017-09-01

In fusion reactors, neutron induced radioactivity strongly depends on the irradiated material. So, a proper selection of structural materials will have been limited the radioactive inventory in a fusion reactor. First-wall and blanket components have high radioactivity concentration due to being the most flux-exposed structures. The main objective of fusion structural material research is the development and selection of materials for reactor components with good thermo-mechanical and physical properties, coupled with low-activation characteristics. Double differential light charged particle emission cross section, which is a fundamental data to determine nuclear heating and material damages in structural fusion material research, for some elements target nuclei have been calculated by the TALYS 1.8 nuclear reaction code at 14-15 MeV neutron incident energy and compared with available experimental data in EXFOR library. Direct, compound and pre-equilibrium reaction contribution have been theoretically calculated and dominant contribution have been determined for each emission of proton, deuteron and alpha particle.

RELAP-7 Code Assessment Plan and Requirement Traceability Matrix

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yoo, Junsoo; Choi, Yong-joon; Smith, Curtis L.

2016-10-01

The RELAP-7, a safety analysis code for nuclear reactor system, is under development at Idaho National Laboratory (INL). Overall, the code development is directed towards leveraging the advancements in computer science technology, numerical solution methods and physical models over the last decades. Recently, INL has also been putting an effort to establish the code assessment plan, which aims to ensure an improved final product quality through the RELAP-7 development process. The ultimate goal of this plan is to propose a suitable way to systematically assess the wide range of software requirements for RELAP-7, including the software design, user interface, andmore » technical requirements, etc. To this end, we first survey the literature (i.e., international/domestic reports, research articles) addressing the desirable features generally required for advanced nuclear system safety analysis codes. In addition, the V&V (verification and validation) efforts as well as the legacy issues of several recently-developed codes (e.g., RELAP5-3D, TRACE V5.0) are investigated. Lastly, this paper outlines the Requirement Traceability Matrix (RTM) for RELAP-7 which can be used to systematically evaluate and identify the code development process and its present capability.« less

PREMOR: a point reactor exposure model computer code for survey analysis of power plant performance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vondy, D.R.

1979-10-01

The PREMOR computer code was written to exploit a simple, two-group point nuclear reactor power plant model for survey analysis. Up to thirteen actinides, fourteen fission products, and one lumped absorber nuclide density are followed over a reactor history. Successive feed batches are accounted for with provision for from one to twenty batches resident. The effect of exposure of each of the batches to the same neutron flux is determined.

Modification of the SAS4A Safety Analysis Code for Integration with the ADAPT Discrete Dynamic Event Tree Framework.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jankovsky, Zachary Kyle; Denman, Matthew R.

It is difficult to assess the consequences of a transient in a sodium-cooled fast reactor (SFR) using traditional probabilistic risk assessment (PRA) methods, as numerous safety-related sys- tems have passive characteristics. Often there is significant dependence on the value of con- tinuous stochastic parameters rather than binary success/failure determinations. One form of dynamic PRA uses a system simulator to represent the progression of a transient, tracking events through time in a discrete dynamic event tree (DDET). In order to function in a DDET environment, a simulator must have characteristics that make it amenable to changing physical parameters midway through themore » analysis. The SAS4A SFR system analysis code did not have these characteristics as received. This report describes the code modifications made to allow dynamic operation as well as the linking to a Sandia DDET driver code. A test case is briefly described to demonstrate the utility of the changes.« less

Verification of RELAP5-3D code in natural circulation loop as function of the initial water inventory

NASA Astrophysics Data System (ADS)

Bertani, C.; Falcone, N.; Bersano, A.; Caramello, M.; Matsushita, T.; De Salve, M.; Panella, B.

2017-11-01

High safety and reliability of advanced nuclear reactors, Generation IV and Small Modular Reactors (SMR), have a crucial role in the acceptance of these new plants design. Among all the possible safety systems, particular efforts are dedicated to the study of passive systems because they rely on simple physical principles like natural circulation, without the need of external energy source to operate. Taking inspiration from the second Decay Heat Removal system (DHR2) of ALFRED, the European Generation IV demonstrator of the fast lead cooled reactor, an experimental facility has been built at the Energy Department of Politecnico di Torino (PROPHET facility) to study single and two-phase flow natural circulation. The facility behavior is simulated using the thermal-hydraulic system code RELAP5-3D, which is widely used in nuclear applications. In this paper, the effect of the initial water inventory on natural circulation is analyzed. The experimental time behaviors of temperatures and pressures are analyzed. The experimental matrix ranges between 69 % and 93%; the influence of the opposite effects related to the increase of the volume available for the expansion and the pressure raise due to phase change is discussed. Simulations of the experimental tests are carried out by using a 1D model at constant heat power and fixed liquid and air mass; the code predictions are compared with experimental results. Two typical responses are observed: subcooled or two phase saturated circulation. The steady state pressure is a strong function of liquid and air mass inventory. The numerical results show that, at low initial liquid mass inventory, the natural circulation is not stable but pulsated.

Dosimetric and microdosimetric analyses for blood exposed to reactor-derived thermal neutrons.

PubMed

Ali, F; Atanackovic, J; Boyer, C; Festarini, A; Kildea, J; Paterson, L C; Rogge, R; Stuart, M; Richardson, R B

2018-06-06

Thermal neutrons are found in reactor, radiotherapy, aircraft, and space environments. The purpose of this study was to characterise the dosimetry and microdosimetry of thermal neutron exposures, using three simulation codes, as a precursor to quantitative radiobiological studies using blood samples. An irradiation line was designed employing a pyrolytic graphite crystal or-alternatively-a super mirror to expose blood samples to thermal neutrons from the National Research Universal reactor to determine radiobiological parameters. The crystal was used when assessing the relative biological effectiveness for dicentric chromosome aberrations, and other biomarkers, in lymphocytes over a low absorbed dose range of 1.2-14 mGy. Higher exposures using a super mirror will allow the additional quantification of mitochondrial responses. The physical size of the thermal neutron fields and their respective wavelength distribution was determined using the McStas Monte Carlo code. Spinning the blood samples produced a spatially uniform absorbed dose as determined from Monte Carlo N-Particle version 6 simulations. The major part (71%) of the total absorbed dose to blood was determined to be from the 14 N(n,p) 14 C reaction and the remainder from the 1 H(n,γ) 2 H reaction. Previous radiobiological experiments at Canadian Nuclear Laboratories involving thermal neutron irradiation of blood yielded a relative biological effectiveness of 26 ± 7. Using the Particle and Heavy Ion Transport Code System, a similar value of ∼19 for the quality factor of thermal neutrons initiating the 14 N(n,p) 14 C reaction in soft tissue was determined by microdosimetric simulations. This calculated quality factor is of similar high value to the experimentally-derived relative biological effectiveness, and indicates the potential of thermal neutrons to induce deleterious health effects in superficial organs such as cataracts of the eye lens.

Particle-in-cell simulations of bounded plasma discharges applied to low pressure high density sources and positive columns

NASA Astrophysics Data System (ADS)

Kawamura, Emi

Particle-in-cell (PIC) simulations of bounded plasma discharges are attractive because the fields and the particle motion can be obtained self-consistently from first principles. Thus, we can accurately model a wide range of nonlocal and kinetic behavior. The only disadvantage is that PIC may be computationally expensive compared to other methods. Fluid codes, for example, may run faster but make assumptions about the bulk plasma velocity distributions and ignore kinetic effects. In Chapter 1, we demonstrate methods of accelerating PIC simulations of bounded plasma discharges. We find that a combination of physical and numerical methods makes run-times for PIC codes much more competitive with other types of codes. In processing plasmas, the ion energy distributions (IEDs) arriving at the wafer target are crucial in determining ion anisotropy and etch rates. The current trend for plasma reactors is towards lower gas pressure and higher plasma density. In Chapter 2, we review and analyze IEDs arriving at the target of low pressure high density rf plasma reactors. In these reactors, the sheath is typically collisionless. We then perform PIC simulations of collisionless rf sheaths and find that the key parameter governing the shape of the TED at the wafer is the ratio of the ion transit time across the sheath over the rf period. Positive columns are the source of illumination in fluorescent mercury-argon lamps. The efficiency of light production increases with decreasing gas pressure and decreasing discharge radius. Most current lamp software is based on the local concept even though low pressure lighting discharges tend to be nonlocal. In Chapter 3, we demonstrate a 1d3v radial PIC model to conduct nonlocal kinetic simulations of low pressure, small radius positive columns. When compared to other available codes, we find that our PIC code makes the least approximations and assumptions and is accurate and stable over a wider parameter range. We analyze the PIC simulation results in detail and find that the radial electron heat flow, which is neglected in local models, plays a major role in maintaining the global power balance. In Chapter 2, we focused on the sheaths of low pressure high density plasma reactors. In Chapter 4, we extend our study to the bulk and presheaths. Typical industrial plasma reactors often use gases with complex chemistries which tend to generate discharges containing negative ions. For high density electronegative plasmas with low gas pressure, we expect Coulomb collisions between positive and negative ions to dominate over collisions between ions and neutrals. We incorporate a Coulomb collision model into our PIC code to study the effect of this ion-ion Coulomb scattering. We find that the Coulomb collisions between the positive and negative ions significantly modify the negative ion flux, density and kinetic energy profiles.

Development of Fuel Shuffling Module for PHISICS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Allan Mabe; Andrea Alfonsi; Cristian Rabiti

2013-06-01

PHISICS (Parallel and Highly Innovative Simulation for the INL Code System) [4] code toolkit has been in development at the Idaho National Laboratory. This package is intended to provide a modern analysis tool for reactor physics investigation. It is designed with the mindset to maximize accuracy for a given availability of computational resources and to give state of the art tools to the modern nuclear engineer. This is obtained by implementing several different algorithms and meshing approaches among which the user will be able to choose, in order to optimize his computational resources and accuracy needs. The software is completelymore » modular in order to simplify the independent development of modules by different teams and future maintenance. The package is coupled with the thermo-hydraulic code RELAP5-3D [3]. In the following the structure of the different PHISICS modules is briefly recalled, focusing on the new shuffling module (SHUFFLE), object of this paper.« less

A Computational Chemistry Database for Semiconductor Processing

NASA Technical Reports Server (NTRS)

Jaffe, R.; Meyyappan, M.; Arnold, J. O. (Technical Monitor)

1998-01-01

The concept of 'virtual reactor' or 'virtual prototyping' has received much attention recently in the semiconductor industry. Commercial codes to simulate thermal CVD and plasma processes have become available to aid in equipment and process design efforts, The virtual prototyping effort would go nowhere if codes do not come with a reliable database of chemical and physical properties of gases involved in semiconductor processing. Commercial code vendors have no capabilities to generate such a database, rather leave the task to the user of finding whatever is needed. While individual investigations of interesting chemical systems continue at Universities, there has not been any large scale effort to create a database. In this presentation, we outline our efforts in this area. Our effort focuses on the following five areas: 1. Thermal CVD reaction mechanism and rate constants. 2. Thermochemical properties. 3. Transport properties.4. Electron-molecule collision cross sections. and 5. Gas-surface interactions.

Response of the first wetted wall of an IFE reactor chamber to the energy release from a direct-drive DT capsule

DOE Office of Scientific and Technical Information (OSTI.GOV)

Medin, Stanislav A.; Basko, Mikhail M.; Orlov, Yurii N.

2012-07-11

Radiation hydrodynamics 1D simulations were performed with two concurrent codes, DEIRA and RAMPHY. The DEIRA code was used for DT capsule implosion and burn, and the RAMPHY code was used for computation of X-ray and fast ions deposition in the first wall liquid film of the reactor chamber. The simulations were run for 740 MJ direct drive DT capsule and Pb thin liquid wall reactor chamber of 10 m diameter. Temporal profiles for DT capsule leaking power of X-rays, neutrons and fast {sup 4}He ions were obtained and spatial profiles of the liquid film flow parameter were computed and analyzed.

Validation Data and Model Development for Fuel Assembly Response to Seismic Loads

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bardet, Philippe; Ricciardi, Guillaume

2016-01-31

Vibrations are inherently present in nuclear reactors, especially in cores and steam generators of pressurized water reactors (PWR). They can have significant effects on local heat transfer and wear and tear in the reactor and often set safety margins. The simulation of these multiphysics phenomena from first principles requires the coupling of several codes, which is one the most challenging tasks in modern computer simulation. Here an ambitious multiphysics multidisciplinary validation campaign is conducted. It relied on an integrated team of experimentalists and code developers to acquire benchmark and validation data for fluid-structure interaction codes. Data are focused on PWRmore » fuel bundle behavior during seismic transients.« less

155. ARAIII Reactor building (ARA608) Details of reactor pit showing ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

155. ARA-III Reactor building (ARA-608) Details of reactor pit showing tray supports and fuel element storage rack. Aerojet-general 880-area/GCRE-608-MS-2. Date: November 1958. Ineel index code no. 063-0608-40-013-102625. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Current and anticipated uses of thermal hydraulic codes at the Japan Atomic Energy Research Institute

DOE Office of Scientific and Technical Information (OSTI.GOV)

Akimoto, Hajime; Kukita; Ohnuki, Akira

1997-07-01

The Japan Atomic Energy Research Institute (JAERI) is conducting several research programs related to thermal-hydraulic and neutronic behavior of light water reactors (LWRs). These include LWR safety research projects, which are conducted in accordance with the Nuclear Safety Commission`s research plan, and reactor engineering projects for the development of innovative reactor designs or core/fuel designs. Thermal-hydraulic and neutronic codes are used for various purposes including experimental analysis, nuclear power plant (NPP) safety analysis, and design assessment.

Current and anticipated uses of thermal-hydraulic codes in Germany

DOE Office of Scientific and Technical Information (OSTI.GOV)

Teschendorff, V.; Sommer, F.; Depisch, F.

1997-07-01

In Germany, one third of the electrical power is generated by nuclear plants. ATHLET and S-RELAP5 are successfully applied for safety analyses of the existing PWR and BWR reactors and possible future reactors, e.g. EPR. Continuous development and assessment of thermal-hydraulic codes are necessary in order to meet present and future needs of licensing organizations, utilities, and vendors. Desired improvements include thermal-hydraulic models, multi-dimensional simulation, computational speed, interfaces to coupled codes, and code architecture. Real-time capability will be essential for application in full-scope simulators. Comprehensive code validation and quantification of uncertainties are prerequisites for future best-estimate analyses.

Thermal Stratification Analysis for Sodium Fast Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schneider, James; Anderson, Mark; Baglietto, Emilio

The sodium fast reactor (SFR) is the most mature reactor concept of all the generation-IV nuclear systems and is a promising reactor design that is currently under development by several organizations. The majority of sodium fast reactor designs utilize a pool type arrangement which incorporates the primary coolant pumps and intermediate heat exchangers within the sodium pool. These components typically protrude into the pool thus reducing the risk and severity of a loss of coolant accidents. To further ensure safe operation under even the most severe transients a more comprehensive understanding of key thermal hydraulic phenomena in this pool ismore » desired. One of the key technology gaps identified for SFR safety is determining the extent and the effects of thermal stratification developing in the pool during postulated accident scenarios such as a protected or unprotected loss of flow incident. In an effort to address these issues, detailed flow models of transient stratification in the pool during an accident can be developed. However, to develop the calculation models, and ensure they can reproduce the underlying physics, highly spatially resolved data is needed. This data can be used in conjunction with advanced computational fluid dynamic calculations to aid in the development of simple reduced dimensional models for systems codes such as SAM and SAS4A/SASSYS-1.« less

IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Strydom, Gerhard; Bostelmann, F.

The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained).more » SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on the HTGR Uncertainty Analysis in Modelling (UAM) be implemented. This CRP is a continuation of the previous IAEA and Organization for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) international activities on Verification and Validation (V&V) of available analytical capabilities for HTGR simulation for design and safety evaluations. Within the framework of these activities different numerical and experimental benchmark problems were performed and insight was gained about specific physics phenomena and the adequacy of analysis methods.« less

NEET Micro-Pocket Fission Detector. Final Project report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Unruh, T.; Rempe, Joy; McGregor, Douglas

2014-09-01

A collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Alternative Energies and Atomic Energy Commission, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), is funded by the Nuclear Energy Enabling Technologies (NEET) program to develop and test Micro-Pocket Fission Detectors (MPFDs), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package. When deployed, these sensors will significantly advance flux detection capabilities for irradiation tests in US Material Test Reactors (MTRs). Ultimately, evaluations may lead to a more compact, more accurate, andmore » longer lifetime flux sensor for critical mock-ups, and high performance reactors, allowing several Department of Energy Office of Nuclear Energy (DOE-NE) programs to obtain higher accuracy/higher resolution data from irradiation tests of candidate new fuels and materials. Specifically, deployment of MPFDs will address several challenges faced in irradiations performed at MTRs: Current fission chamber technologies do not offer the ability to measure fast flux, thermal flux and temperature within a single compact probe; MPFDs offer this option. MPFD construction is very different than current fission chamber construction; the use of high temperature materials allow MPFDs to be specifically tailored to survive harsh conditions encountered in-core of high performance MTRs. The higher accuracy, high fidelity data available from the compact MPFD will significantly enhance efforts to validate new high-fidelity reactor physics codes and new multi-scale, multi-physics codes. MPFDs can be built with variable sensitivities to survive the lifetime of an experiment or fuel assembly in some MTRs, allowing for more efficient and cost effective power monitoring. The small size of the MPFDs allows multiple sensors to be deployed, offering the potential to accurately measure the flux and temperature profiles in the reactor. This report summarizes the status at the end of year two of this three year project. As documented in this report, all planned accomplishments for developing this unique new, compact, multipurpose sensor have been completed.« less

10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... nonpower reactors. 73.60 Section 73.60 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION... requirements for physical protection at nonpower reactors. Each nonpower reactor licensee who, pursuant to the... nonpower reactors licensed to operate at or above a power level of 2 megawatts thermal. [38 FR 35430, Dec...

10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... nonpower reactors. 73.60 Section 73.60 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION... requirements for physical protection at nonpower reactors. Each nonpower reactor licensee who, pursuant to the... nonpower reactors licensed to operate at or above a power level of 2 megawatts thermal. [38 FR 35430, Dec...

Computational fluid dynamics assessment: Volume 1, Computer simulations of the METC (Morgantown Energy Technology Center) entrained-flow gasifier: Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Celik, I.; Chattree, M.

1988-07-01

An assessment of the theoretical and numerical aspects of the computer code, PCGC-2, is made; and the results of the application of this code to the Morgantown Energy Technology Center (METC) advanced gasification facility entrained-flow reactor, ''the gasifier,'' are presented. PCGC-2 is a code suitable for simulating pulverized coal combustion or gasification under axisymmetric (two-dimensional) flow conditions. The governing equations for the gas and particulate phase have been reviewed. The numerical procedure and the related programming difficulties have been elucidated. A single-particle model similar to the one used in PCGC-2 has been developed, programmed, and applied to some simple situationsmore » in order to gain insight to the physics of coal particle heat-up, devolatilization, and char oxidation processes. PCGC-2 was applied to the METC entrained-flow gasifier to study numerically the flash pyrolysis of coal, and gasification of coal with steam or carbon dioxide. The results from the simulations are compared with measurements. The gas and particle residence times, particle temperature, and mass component history were also calculated and the results were analyzed. The results provide useful information for understanding the fundamentals of coal gasification and for assessment of experimental results performed using the reactor considered. 69 refs., 35 figs., 23 tabs.« less

Full core analysis of IRIS reactor by using MCNPX.

PubMed

Amin, E A; Bashter, I I; Hassan, Nabil M; Mustafa, S S

2016-07-01

This paper describes neutronic analysis for fresh fuelled IRIS (International Reactor Innovative and Secure) reactor by MCNPX code. The analysis included criticality calculations, radial power and axial power distribution, nuclear peaking factor and axial offset percent at the beginning of fuel cycle. The effective multiplication factor obtained by MCNPX code is compared with previous calculations by HELIOS/NESTLE, CASMO/SIMULATE, modified CORD-2 nodal calculations and SAS2H/KENO-V code systems. It is found that k-eff value obtained by MCNPX is closer to CORD-2 value. The radial and axial powers are compared with other published results carried out using SAS2H/KENO-V code. Moreover, the WIMS-D5 code is used for studying the effect of enriched boron in form of ZrB2 on the effective multiplication factor (K-eff) of the fuel pin. In this part of calculation, K-eff is calculated at different concentrations of Boron-10 in mg/cm at different stages of burnup of unit cell. The results of this part are compared with published results performed by HELIOS code. Copyright © 2016 Elsevier Ltd. All rights reserved.

Exposure calculation code module for reactor core analysis: BURNER

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vondy, D.R.; Cunningham, G.W.

1979-02-01

The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also providesmore » user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.« less

Nuclear Reactor Physics

NASA Astrophysics Data System (ADS)

Stacey, Weston M.

2001-02-01

An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

Hybrid Reduced Order Modeling Algorithms for Reactor Physics Calculations

NASA Astrophysics Data System (ADS)

Bang, Youngsuk

Reduced order modeling (ROM) has been recognized as an indispensable approach when the engineering analysis requires many executions of high fidelity simulation codes. Examples of such engineering analyses in nuclear reactor core calculations, representing the focus of this dissertation, include the functionalization of the homogenized few-group cross-sections in terms of the various core conditions, e.g. burn-up, fuel enrichment, temperature, etc. This is done via assembly calculations which are executed many times to generate the required functionalization for use in the downstream core calculations. Other examples are sensitivity analysis used to determine important core attribute variations due to input parameter variations, and uncertainty quantification employed to estimate core attribute uncertainties originating from input parameter uncertainties. ROM constructs a surrogate model with quantifiable accuracy which can replace the original code for subsequent engineering analysis calculations. This is achieved by reducing the effective dimensionality of the input parameter, the state variable, or the output response spaces, by projection onto the so-called active subspaces. Confining the variations to the active subspace allows one to construct an ROM model of reduced complexity which can be solved more efficiently. This dissertation introduces a new algorithm to render reduction with the reduction errors bounded based on a user-defined error tolerance which represents the main challenge of existing ROM techniques. Bounding the error is the key to ensuring that the constructed ROM models are robust for all possible applications. Providing such error bounds represents one of the algorithmic contributions of this dissertation to the ROM state-of-the-art. Recognizing that ROM techniques have been developed to render reduction at different levels, e.g. the input parameter space, the state space, and the response space, this dissertation offers a set of novel hybrid ROM algorithms which can be readily integrated into existing methods and offer higher computational efficiency and defendable accuracy of the reduced models. For example, the snapshots ROM algorithm is hybridized with the range finding algorithm to render reduction in the state space, e.g. the flux in reactor calculations. In another implementation, the perturbation theory used to calculate first order derivatives of responses with respect to parameters is hybridized with a forward sensitivity analysis approach to render reduction in the parameter space. Reduction at the state and parameter spaces can be combined to render further reduction at the interface between different physics codes in a multi-physics model with the accuracy quantified in a similar manner to the single physics case. Although the proposed algorithms are generic in nature, we focus here on radiation transport models used in support of the design and analysis of nuclear reactor cores. In particular, we focus on replacing the traditional assembly calculations by ROM models to facilitate the generation of homogenized cross-sections for downstream core calculations. The implication is that assembly calculations could be done instantaneously therefore precluding the need for the expensive evaluation of the few-group cross-sections for all possible core conditions. Given the generic natures of the algorithms, we make an effort to introduce the material in a general form to allow non-nuclear engineers to benefit from this work.

Investigation of two-phase phenomena occurring within moisture separator reheater high-level reactor trips at the Maanshan nuclear power plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ferng, Y.M.; Liao, L.Y.

1996-01-01

During the operating history of the Maanshan nuclear power plant (MNPP), five reactor trips have occurred as a result of the moisture separator reheater (MSR) high-level signal. These MSR high-level reactor trips have been a very serious concern, especially during the startup period of MNPP. Consequently, studying the physical phenomena of this particular event is worthwhile, and analytical work is performed using the RELAP5/MOD3 code to investigate the thermal-hydraulic phenomena of two-phase behaviors occurring within the MSR high-level reactor trips. The analytical model is first assessed against the experimental data obtained from several test loops. The same model can thenmore » be applied with confidence to the study of this topic. According to the present calculated results, the phenomena of liquid droplet accumulation ad residual liquid blowing in the horizontal section of cross-under-lines can be modeled. In addition, the present model can also predict the different increasing rates of inlet steam flow rate affecting the liquid accumulation within the cross-under-lines. The calculated conclusion is confirmed by the revised startup procedure of MNPP.« less

Warthog: Progress on Coupling BISON and PROTEUS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hart, Shane W.D.

The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program from the Office of Nuclear Energy at the Department of Energy (DOE) provides a robust toolkit for modeling and simulation of current and future advanced nuclear reactor designs. This toolkit provides these technologies organized across product lines, with two divisions targeted at fuels and end-to-end reactor modeling, and a third for integration, coupling, and high-level workflow management. The Fuels Product Line (FPL) and the Reactor Product Line (RPL) provide advanced computational technologies that serve each respective field effectively. There is currently a lack of integration between the product lines, impeding futuremore » improvements of simulation solution fidelity. In order to mix and match tools across the product lines, a new application called Warthog was produced. Warthog is built on the Multi-physics Object-Oriented Simulation Environment (MOOSE) framework developed at Idaho National Laboratory (INL). This report details the continuing efforts to provide the Integration Product Line (IPL) with interoperability using the Warthog code. Currently, this application strives to couple the BISON fuel performance application from the FPL using the PROTEUS Core Neutronics application from the RPL. Warthog leverages as much prior work from the NEAMS program as possible, enabling interoperability between the independently developed MOOSE and SHARP frameworks, and the libMesh and MOAB mesh data formats. Previous work performed on Warthog allowed it to couple a pin cell between the two codes. However, as the temperature changed due to the BISON calculation, the cross sections were not recalculated, leading to errors as the temperature got further away from the initial conditions. XSProc from the SCALE code suite was used to calculate the cross sections as needed. The remainder of this report discusses the changes to Warthog to allow for the implementation of XSProc as an external code. It also discusses the changes made to Warthog to allow it to fit more cleanly into the MultiApp syntax of the MOOSE framework. The capabilities, design, and limitations of Warthog will be described, in addition to some of the test cases that were used to demonstrate the code. Future plans for Warthog will be discussed, including continuation of the modifications to the input and coupling to other SHARP codes such as Nek5000.« less

78 FR 20959 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-08

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR... Committee on Reactor Safeguards. [FR Doc. 2013-08131 Filed 4-5-13; 8:45 am] BILLING CODE 7590-01-P ...

A multi-physics analysis for the actuation of the SSS in opal reactor

NASA Astrophysics Data System (ADS)

Ferraro, Diego; Alberto, Patricio; Villarino, Eduardo; Doval, Alicia

2018-05-01

OPAL is a 20 MWth multi-purpose open-pool type Research Reactor located at Lucas Heights, Australia. It was designed, built and commissioned by INVAP between 2000 and 2006 and it has been operated by the Australia Nuclear Science and Technology Organization (ANSTO) showing a very good overall performance. On November 2016, OPAL reached 10 years of continuous operation, becoming one of the most reliable and available in its kind worldwide, with an unbeaten record of being fully operational 307 days a year. One of the enhanced safety features present in this state-of-art reactor is the availability of an independent, diverse and redundant Second Shutdown System (SSS), which consists in the drainage of the heavy water reflector contained in the Reflector Vessel. As far as high quality experimental data is available from reactor commissioning and operation stages and even from early component design validation stages, several models both regarding neutronic and thermo-hydraulic approaches have been developed during recent years using advanced calculations tools and the novel capabilities to couple them. These advanced models were developed in order to assess the capability of such codes to simulate and predict complex behaviours and develop highly detail analysis. In this framework, INVAP developed a three-dimensional CFD model that represents the detailed hydraulic behaviour of the Second Shutdown System for an actuation scenario, where the heavy water drainage 3D temporal profiles inside the Reflector Vessel can be obtained. This model was validated, comparing the computational results with experimental measurements performed in a real-size physical model built by INVAP during early OPAL design engineering stages. Furthermore, detailed 3D Serpent Monte Carlo models are also available, which have been already validated with experimental data from reactor commissioning and operating cycles. In the present work the neutronic and thermohydraulic models, available for OPAL reactor, are coupled by means of a shared unstructured mesh geometry definition of relevant zones inside the Reflector Vessel. Several scenarios, both regarding coupled and uncoupled neutronic & thermohydraulic behavior, are presented and analyzed, showing the capabilities to develop and manage advanced modelling that allows to predict multi-physics variables observed when an in-depth performance analysis of a Research Reactor like OPAL is carried out.

SHARP Multiphysics Tutorials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yu, Y. Q.; Shemon, E. R.; Mahadevan, Vijay S.

SHARP, developed under the NEAMS Reactor Product Line, is an advanced modeling and simulation toolkit for the analysis of advanced nuclear reactors. SHARP is comprised of three physics modules currently including neutronics, thermal hydraulics, and structural mechanics. SHARP empowers designers to produce accurate results for modeling physical phenomena that have been identified as important for nuclear reactor analysis. SHARP can use existing physics codes and take advantage of existing infrastructure capabilities in the MOAB framework and the coupling driver/solver library, the Coupled Physics Environment (CouPE), which utilizes the widely used, scalable PETSc library. This report aims at identifying the coupled-physicsmore » simulation capability of SHARP by introducing the demonstration example called sahex in advance of the SHARP release expected by Mar 2016. sahex consists of 6 fuel pins with cladding, 1 control rod, sodium coolant and an outer duct wall that encloses all the other components. This example is carefully chosen to demonstrate the proof of concept for solving more complex demonstration examples such as EBR II assembly and ABTR full core. The workflow of preparing the input files, running the case and analyzing the results is demonstrated in this report. Moreover, an extension of the sahex model called sahex_core, which adds six homogenized neighboring assemblies to the full heterogeneous sahex model, is presented to test homogenization capabilities in both Nek5000 and PROTEUS. Some primary information on the configuration and build aspects for the SHARP toolkit, which includes capability to auto-download dependencies and configure/install with optimal flags in an architecture-aware fashion, is also covered by this report. A step-by-step instruction is provided to help users to create their cases. Details on these processes will be provided in the SHARP user manual that will accompany the first release.« less

Criticality calculations of the Very High Temperature reactor Critical Assembly benchmark with Serpent and SCALE/KENO-VI

DOE PAGES

Bostelmann, Friederike; Hammer, Hans R.; Ortensi, Javier; ...

2015-12-30

Within the framework of the IAEA Coordinated Research Project on HTGR Uncertainty Analysis in Modeling, criticality calculations of the Very High Temperature Critical Assembly experiment were performed as the validation reference to the prismatic MHTGR-350 lattice calculations. Criticality measurements performed at several temperature points at this Japanese graphite-moderated facility were recently included in the International Handbook of Evaluated Reactor Physics Benchmark Experiments, and represent one of the few data sets available for the validation of HTGR lattice physics. Here, this work compares VHTRC criticality simulations utilizing the Monte Carlo codes Serpent and SCALE/KENO-VI. Reasonable agreement was found between Serpent andmore » KENO-VI, but only the use of the latest ENDF cross section library release, namely the ENDF/B-VII.1 library, led to an improved match with the measured data. Furthermore, the fourth beta release of SCALE 6.2/KENO-VI showed significant improvements from the current SCALE 6.1.2 version, compared to the experimental values and Serpent.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ritchie, L.T.; Johnson, J.D.; Blond, R.M.

The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems.

TEMPEST code simulations of hydrogen distribution in reactor containment structures. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trent, D.S.; Eyler, L.L.

The mass transport version of the TEMPEST computer code was used to simulate hydrogen distribution in geometric configurations relevant to reactor containment structures. Predicted results of Battelle-Frankfurt hydrogen distribution tests 1 to 6, and 12 are presented. Agreement between predictions and experimental data is good. Best agreement is obtained using the k-epsilon turbulence model in TEMPEST in flow cases where turbulent diffusion and stable stratification are dominant mechanisms affecting transport. The code's general analysis capabilities are summarized.

Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2012

DOE Office of Scientific and Technical Information (OSTI.GOV)

David W. Nigg, Principal Investigator; Kevin A. Steuhm, Project Manager

Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to properly verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Updatemore » Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the next anticipated ATR Core Internals Changeout (CIC) in the 2014-2015 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its third full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL under various licensing arrangements. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purpose. Of particular importance, a set of as-run core depletion HELIOS calculations for all ATR cycles since August 2009, Cycle 145A through Cycle 151B, was successfully completed during 2012. This major effort supported a decision late in the year to proceed with the phased incorporation of the HELIOS methodology into the ATR Core Safety Analysis Package (CSAP) preparation process, in parallel with the established PDQ-based methodology, beginning late in Fiscal Year 2012. Acquisition of the advanced SERPENT (VTT-Finland) and MC21 (DOE-NR) Monte Carlo stochastic neutronics simulation codes was also initiated during the year and some initial applications of SERPENT to ATRC experiment analysis were demonstrated. These two new codes will offer significant additional capability, including the possibility of full-3D Monte Carlo fuel management support capabilities for the ATR at some point in the future. Finally, a capability for rigorous sensitivity analysis and uncertainty quantification based on the TSUNAMI system has been implemented and initial computational results have been obtained. This capability will have many applications as a tool for understanding the margins of uncertainty in the new models as well as for validation experiment design and interpretation.« less

Three-dimensional time-dependent STAR reactor kinetics analyses coupled with RETRAN and MCPWR system response

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.

1989-11-01

The operation of a nuclear power plant must be regularly supported by various reactor dynamics and thermal-hydraulic analyses, which may include final safety analysis report (FSAR) design-basis calculations, and conservative and best-estimate analyses. The development and improvement of computer codes and analysis methodologies provide many advantages, including the ability to evaluate the effect of modeling simplifications and assumptions made in previous reactor kinetics and thermal-hydraulic calculations. This paper describes the results of using the RETRAN, MCPWR, and STAR codes in a tandem, predictive-corrective manner for three pressurized water reactor (PWR) transients: (a) loss of feedwater (LOF) anticipated transient without scrammore » (ATWS), (b) station blackout ATWS, and (c) loss of total reactor coolant system (RCS) flow with a scram.« less

Optimization of 3D Field Design

NASA Astrophysics Data System (ADS)

Logan, Nikolas; Zhu, Caoxiang

2017-10-01

Recent progress in 3D tokamak modeling is now leveraged to create a conceptual design of new external 3D field coils for the DIII-D tokamak. Using the IPEC dominant mode as a target spectrum, the Finding Optimized Coils Using Space-curves (FOCUS) code optimizes the currents and 3D geometry of multiple coils to maximize the total set's resonant coupling. The optimized coils are individually distorted in space, creating toroidal ``arrays'' containing a variety of shapes that often wrap around a significant poloidal extent of the machine. The generalized perturbed equilibrium code (GPEC) is used to determine optimally efficient spectra for driving total, core, and edge neoclassical toroidal viscosity (NTV) torque and these too provide targets for the optimization of 3D coil designs. These conceptual designs represent a fundamentally new approach to 3D coil design for tokamaks targeting desired plasma physics phenomena. Optimized coil sets based on plasma response theory will be relevant to designs for future reactors or on any active machine. External coils, in particular, must be optimized for reliable and efficient fusion reactor designs. Work supported by the US Department of Energy under DE-AC02-09CH11466.

CESAR5.3: Isotopic depletion for Research and Testing Reactor decommissioning

NASA Astrophysics Data System (ADS)

Ritter, Guillaume; Eschbach, Romain; Girieud, Richard; Soulard, Maxime

2018-05-01

CESAR stands in French for "simplified depletion applied to reprocessing". The current version is now number 5.3 as it started 30 years ago from a long lasting cooperation with ORANO, co-owner of the code with CEA. This computer code can characterize several types of nuclear fuel assemblies, from the most regular PWR power plants to the most unexpected gas cooled and graphite moderated old timer research facility. Each type of fuel can also include numerous ranges of compositions like UOX, MOX, LEU or HEU. Such versatility comes from a broad catalog of cross section libraries, each corresponding to a specific reactor and fuel matrix design. CESAR goes beyond fuel characterization and can also provide an evaluation of structural materials activation. The cross-sections libraries are generated using the most refined assembly or core level transport code calculation schemes (CEA APOLLO2 or ERANOS), based on the European JEFF3.1.1 nuclear data base. Each new CESAR self shielded cross section library benefits all most recent CEA recommendations as for deterministic physics options. Resulting cross sections are organized as a function of burn up and initial fuel enrichment which allows to condensate this costly process into a series of Legendre polynomials. The final outcome is a fast, accurate and compact CESAR cross section library. Each library is fully validated, against a stochastic transport code (CEA TRIPOLI 4) if needed and against a reference depletion code (CEA DARWIN). Using CESAR does not require any of the neutron physics expertise implemented into cross section libraries generation. It is based on top quality nuclear data (JEFF3.1.1 for ˜400 isotopes) and includes up to date Bateman equation solving algorithms. However, defining a CESAR computation case can be very straightforward. Most results are only 3 steps away from any beginner's ambition: Initial composition, in core depletion and pool decay scenario. On top of a simple utilization architecture, CESAR includes a portable Graphical User Interface which can be broadly deployed in R&D or industrial facilities. Aging facilities currently face decommissioning and dismantling issues. This way to the end of the nuclear fuel cycle requires a careful assessment of source terms in the fuel, core structures and all parts of a facility that must be disposed of with "industrial nuclear" constraints. In that perspective, several CESAR cross section libraries were constructed for early CEA Research and Testing Reactors (RTR's). The aim of this paper is to describe how CESAR operates and how it can be used to help these facilities care for waste disposal, nuclear materials transport or basic safety cases. The test case will be based on the PHEBUS Facility located at CEA - Cadarache.

GLSENS: A Generalized Extension of LSENS Including Global Reactions and Added Sensitivity Analysis for the Perfectly Stirred Reactor

NASA Technical Reports Server (NTRS)

Bittker, David A.

1996-01-01

A generalized version of the NASA Lewis general kinetics code, LSENS, is described. The new code allows the use of global reactions as well as molecular processes in a chemical mechanism. The code also incorporates the capability of performing sensitivity analysis calculations for a perfectly stirred reactor rapidly and conveniently at the same time that the main kinetics calculations are being done. The GLSENS code has been extensively tested and has been found to be accurate and efficient. Nine example problems are presented and complete user instructions are given for the new capabilities. This report is to be used in conjunction with the documentation for the original LSENS code.

FY2012 summary of tasks completed on PROTEUS-thermal work.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, C.H.; Smith, M.A.

2012-06-06

PROTEUS is a suite of the neutronics codes, both old and new, that can be used within the SHARP codes being developed under the NEAMS program. Discussion here is focused on updates and verification and validation activities of the SHARP neutronics code, DeCART, for application to thermal reactor analysis. As part of the development of SHARP tools, the different versions of the DeCART code created for PWR, BWR, and VHTR analysis were integrated. Verification and validation tests for the integrated version were started, and the generation of cross section libraries based on the subgroup method was revisited for the targetedmore » reactor types. The DeCART code has been reorganized in preparation for an efficient integration of the different versions for PWR, BWR, and VHTR analysis. In DeCART, the old-fashioned common blocks and header files have been replaced by advanced memory structures. However, the changing of variable names was minimized in order to limit problems with the code integration. Since the remaining stability problems of DeCART were mostly caused by the CMFD methodology and modules, significant work was performed to determine whether they could be replaced by more stable methods and routines. The cross section library is a key element to obtain accurate solutions. Thus, the procedure for generating cross section libraries was revisited to provide libraries tailored for the targeted reactor types. To improve accuracy in the cross section library, an attempt was made to replace the CENTRM code by the MCNP Monte Carlo code as a tool obtaining reference resonance integrals. The use of the Monte Carlo code allows us to minimize problems or approximations that CENTRM introduces since the accuracy of the subgroup data is limited by that of the reference solutions. The use of MCNP requires an additional set of libraries without resonance cross sections so that reference calculations can be performed for a unit cell in which only one isotope of interest includes resonance cross sections, among the isotopes in the composition. The OECD MHTGR-350 benchmark core was simulated using DeCART as initial focus of the verification/validation efforts. Among the benchmark problems, Exercise 1 of Phase 1 is a steady-state benchmark case for the neutronics calculation for which block-wise cross sections were provided in 26 energy groups. This type of problem was designed for a homogenized geometry solver like DIF3D rather than the high-fidelity code DeCART. Instead of the homogenized block cross sections given in the benchmark, the VHTR-specific 238-group ENDF/B-VII.0 library of DeCART was directly used for preliminary calculations. Initial results showed that the multiplication factors of a fuel pin and a fuel block with or without a control rod hole were off by 6, -362, and -183 pcm Dk from comparable MCNP solutions, respectively. The 2-D and 3-D one-third core calculations were also conducted for the all-rods-out (ARO) and all-rods-in (ARI) configurations, producing reasonable results. Figure 1 illustrates the intermediate (1.5 eV - 17 keV) and thermal (below 1.5 eV) group flux distributions. As seen from VHTR cores with annular fuels, the intermediate group fluxes are relatively high in the fuel region, but the thermal group fluxes are higher in the inner and outer graphite reflector regions than in the fuel region. To support the current project, a new three-year I-NERI collaboration involving ANL and KAERI was started in November 2011, focused on performing in-depth verification and validation of high-fidelity multi-physics simulation codes for LWR and VHTR. The work scope includes generating improved cross section libraries for the targeted reactor types, developing benchmark models for verification and validation of the neutronics code with or without thermo-fluid feedback, and performing detailed comparisons of predicted reactor parameters against both Monte Carlo solutions and experimental measurements. The following list summarizes the work conducted so far for PROTEUS-Thermal Tasks: Unification of different versions of DeCART was initiated, and at the same time code modernization was conducted to make code unification efficient; (2) Regeneration of cross section libraries was attempted for the targeted reactor types, and the procedure for generating cross section libraries was updated by replacing CENTRM with MCNP for reference resonance integrals; (3) The MHTGR-350 benchmark core was simulated using DeCART with VHTR-specific 238-group ENDF/B-VII.0 library, and MCNP calculations were performed for comparison; and (4) Benchmark problems for PWR and BWR analysis were prepared for the DeCART verification/validation effort. In the coming months, the work listed above will be completed. Cross section libraries will be generated with optimized group structures for specific reactor types.« less

Validation of light water reactor calculation methods and JEF-1-based data libraries by TRX and BAPL critical experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paratte, J.M.; Pelloni, S.; Grimm, P.

1991-04-01

This paper analyzes the capability of various code systems and JEF-1-based nuclear data libraries to compute light water reactor lattices by comparing calculations with results from thermal reactor benchmark experiments TRX and BAPL and with previously published values. With the JEF-1 evaluation, eigenvalues are generally well predicted within 8 mk (1 mk = 0.001) or less by all code systems, and all methods give reasonable results for the measured reaction rate ratios within, or not too far from, the experimental uncertainty.

NATCRCTR: One-dimensional thermal-hydraulics analysis code for natural-circulation TRIGA reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.; Rubinaccio, G.

1996-12-31

The Pennsylvania State University nuclear engineering department is evaluating the upgrade of the Reed College (Portland, Oregon) TRIGA reactor from 250 kW to 1 MW in two areas: thermal-hydraulics and steady-state neutronics analysis. This analysis was initiated as a cooperative effort between Penn State and Reed College as a training project for two International Atomic Energy Agency (IAEA) fellows from Ghana. The two Ghanaian IAEA fellows were assisted by G. Rubinaccio, an undergraduate, who undertook the task of writing the new computer programs for the thermal-hydraulic and physics evaluation as a three-credit special design project course. The Reed College TRIGA,more » which has a fixed graphite radial reflector, is cooled by natural circulation, without external cross-flow; whereas, the Penn State Breazeale Reactor has significant crossflow into its sides. To model the Reed TRIGA, the NATCRCTR program has been developed from first principles using the following assumptions: 1. The core is surrounded by the fixed reflector structure, which acts as a one-dimensional channel. 2. The core inlet temperature distribution is constant at the core bottom. 3. The axial heat flux distribution is a chopped cosine shape. 4. The heat transfer in the fuel is primarily in the radial directions. 5. A small gap between the fuel and cladding exists. The NATCRCTR code is used to find the peak centerline fuel, gap, and cladding surface temperatures, based on assumed flux and engineering peaking factors.« less

Documentation of probabilistic fracture mechanics codes used for reactor pressure vessels subjected to pressurized thermal shock loading: Parts 1 and 2. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Balkey, K.; Witt, F.J.; Bishop, B.A.

1995-06-01

Significant attention has been focused on the issue of reactor vessel pressurized thermal shock (PTS) for many years. Pressurized thermal shock transient events are characterized by a rapid cooldown at potentially high pressure levels that could lead to a reactor vessel integrity concern for some pressurized water reactors. As a result of regulatory and industry efforts in the early 1980`s, a probabilistic risk assessment methodology has been established to address this concern. Probabilistic fracture mechanics analyses are performed as part of this methodology to determine conditional probability of significant flaw extension for given pressurized thermal shock events. While recent industrymore » efforts are underway to benchmark probabilistic fracture mechanics computer codes that are currently used by the nuclear industry, Part I of this report describes the comparison of two independent computer codes used at the time of the development of the original U.S. Nuclear Regulatory Commission (NRC) pressurized thermal shock rule. The work that was originally performed in 1982 and 1983 to compare the U.S. NRC - VISA and Westinghouse (W) - PFM computer codes has been documented and is provided in Part I of this report. Part II of this report describes the results of more recent industry efforts to benchmark PFM computer codes used by the nuclear industry. This study was conducted as part of the USNRC-EPRI Coordinated Research Program for reviewing the technical basis for pressurized thermal shock (PTS) analyses of the reactor pressure vessel. The work focused on the probabilistic fracture mechanics (PFM) analysis codes and methods used to perform the PTS calculations. An in-depth review of the methodologies was performed to verify the accuracy and adequacy of the various different codes. The review was structured around a series of benchmark sample problems to provide a specific context for discussion and examination of the fracture mechanics methodology.« less

Fuel management optimization using genetic algorithms and expert knowledge

DOE Office of Scientific and Technical Information (OSTI.GOV)

DeChaine, M.D.; Feltus, M.A.

1996-09-01

The CIGARO fuel management optimization code based on genetic algorithms is described and tested. The test problem optimized the core lifetime for a pressurized water reactor with a penalty function constraint on the peak normalized power. A bit-string genotype encoded the loading patterns, and genotype bias was reduced with additional bits. Expert knowledge about fuel management was incorporated into the genetic algorithm. Regional crossover exchanged physically adjacent fuel assemblies and improved the optimization slightly. Biasing the initial population toward a known priority table significantly improved the optimization.

156. ARAIII Reactor building (ARA608) Electrical and control details of ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

156. ARA-III Reactor building (ARA-608) Electrical and control details of mobile work bridge over reactor and pipiing pits. Aerojet-general 880-area/GCRE-608-E-6. Date: November 1958. Ineel index code no. 063-0608-10-013-102621. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Calculation of the Phenix end-of-life test 'Control Rod Withdrawal' with the ERANOS code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tiberi, V.

2012-07-01

The Inst. of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to French public authorities. As such, IRSN is in charge of safety assessment of operating and under construction reactors, as well as future projects. In this framework, one current objective of IRSN is to evaluate the ability and accuracy of numerical tools to foresee consequences of accidents. Neutronic studies step in the safety assessment from different points of view among which the core design and its protection system. They are necessary to evaluate the core behavior in case of accident in order to assess the integrity ofmore » the first barrier and the absence of a prompt criticality risk. To reach this objective one main physical quantity has to be evaluated accurately: the neutronic power distribution in core during whole reactor lifetime. Phenix end of life tests, carried out in 2009, aim at increasing the experience feedback on sodium cooled fast reactors. These experiments have been done in the framework of the development of the 4. generation of nuclear reactors. Ten tests have been carried out: 6 on neutronic and fuel aspects, 2 on thermal hydraulics and 2 for the emergency shutdown. Two of them have been chosen for an international exercise on thermal hydraulics and neutronics in the frame of an IAEA Coordinated Research Project. Concerning neutronics, the Control Rod Withdrawal test is relevant for safety because it allows evaluating the capability of calculation tools to compute the radial power distribution on fast reactors core configurations in which the flux field is very deformed. IRSN participated to this benchmark with the ERANOS code developed by CEA for fast reactors studies. This paper presents the results obtained in the framework of the benchmark activity. A relatively good agreement was found with available measures considering the approximations done in the modeling. The work underlines the importance of burn-up calculations in order to have a fine core concentrations mesh for the calculation of the power distribution. (authors)« less

Neutron transport analysis for nuclear reactor design

DOEpatents

Vujic, Jasmina L.

1993-01-01

Replacing regular mesh-dependent ray tracing modules in a collision/transfer probability (CTP) code with a ray tracing module based upon combinatorial geometry of a modified geometrical module (GMC) provides a general geometry transfer theory code in two dimensions (2D) for analyzing nuclear reactor design and control. The primary modification of the GMC module involves generation of a fixed inner frame and a rotating outer frame, where the inner frame contains all reactor regions of interest, e.g., part of a reactor assembly, an assembly, or several assemblies, and the outer frame, with a set of parallel equidistant rays (lines) attached to it, rotates around the inner frame. The modified GMC module allows for determining for each parallel ray (line), the intersections with zone boundaries, the path length between the intersections, the total number of zones on a track, the zone and medium numbers, and the intersections with the outer surface, which parameters may be used in the CTP code to calculate collision/transfer probability and cross-section values.

Neutron transport analysis for nuclear reactor design

DOEpatents

Vujic, J.L.

1993-11-30

Replacing regular mesh-dependent ray tracing modules in a collision/transfer probability (CTP) code with a ray tracing module based upon combinatorial geometry of a modified geometrical module (GMC) provides a general geometry transfer theory code in two dimensions (2D) for analyzing nuclear reactor design and control. The primary modification of the GMC module involves generation of a fixed inner frame and a rotating outer frame, where the inner frame contains all reactor regions of interest, e.g., part of a reactor assembly, an assembly, or several assemblies, and the outer frame, with a set of parallel equidistant rays (lines) attached to it, rotates around the inner frame. The modified GMC module allows for determining for each parallel ray (line), the intersections with zone boundaries, the path length between the intersections, the total number of zones on a track, the zone and medium numbers, and the intersections with the outer surface, which parameters may be used in the CTP code to calculate collision/transfer probability and cross-section values. 28 figures.

IAEA coordinated research project on thermal-hydraulics of Supercritical Water-Cooled Reactors (SCWRs)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yamada, K.; Aksan, S. N.

The Supercritical Water-Cooled Reactor (SCWR) is an innovative water-cooled reactor concept, which uses supercritical pressure water as reactor coolant. It has been attracting interest of many researchers in various countries mainly due to its benefits of high thermal efficiency and simple primary systems, resulting in low capital cost. The IAEA started in 2008 a Coordinated Research Project (CRP) on Thermal-Hydraulics of SCWRs as a forum to foster the exchange of technical information and international collaboration in research and development. This paper summarizes the activities and current status of the CRP, as well as major progress achieved to date. At present,more » 15 institutions closely collaborate in several tasks. Some organizations have been conducting thermal-hydraulics experiments and analysing the data, and others have been participating in code-to-test and/or code-to-code benchmark exercises. The expected outputs of the CRP are also discussed. Finally, the paper introduces several IAEA activities relating to or arising from the CRP. (authors)« less

Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1992-04-01

This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Requirements for physical protection of irradiated reactor... Requirements for physical protection of irradiated reactor fuel in transit. (a) Performance objectives. (1... of irradiated reactor fuel in excess of 100 grams in net weight of irradiated fuel, exclusive of...

10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Requirements for physical protection of irradiated reactor... Requirements for physical protection of irradiated reactor fuel in transit. (a) Performance objectives. (1... of irradiated reactor fuel in excess of 100 grams in net weight of irradiated fuel, exclusive of...

User's manual for COAST 4: a code for costing and sizing tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sink, D. A.; Iwinski, E. M.

1979-09-01

The purpose of this report is to document the computer program COAST 4 for the user/analyst. COAST, COst And Size Tokamak reactors, provides complete and self-consistent size models for the engineering features of D-T burning tokamak reactors and associated facilities involving a continuum of performance including highly beam driven through ignited plasma devices. TNS (The Next Step) devices with no tritium breeding or electrical power production are handled as well as power producing and fissile producing fusion-fission hybrid reactors. The code has been normalized with a TFTR calculation which is consistent with cost, size, and performance data published in themore » conceptual design report for that device. Information on code development, computer implementation and detailed user instructions are included in the text.« less

Accuracy and convergence of coupled finite-volume/Monte Carlo codes for plasma edge simulations of nuclear fusion reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ghoos, K., E-mail: kristel.ghoos@kuleuven.be; Dekeyser, W.; Samaey, G.

2016-10-01

The plasma and neutral transport in the plasma edge of a nuclear fusion reactor is usually simulated using coupled finite volume (FV)/Monte Carlo (MC) codes. However, under conditions of future reactors like ITER and DEMO, convergence issues become apparent. This paper examines the convergence behaviour and the numerical error contributions with a simplified FV/MC model for three coupling techniques: Correlated Sampling, Random Noise and Robbins Monro. Also, practical procedures to estimate the errors in complex codes are proposed. Moreover, first results with more complex models show that an order of magnitude speedup can be achieved without any loss in accuracymore » by making use of averaging in the Random Noise coupling technique.« less

Comparison of ENDF/B-VII.1 and JEFF-3.2 in VVER-1000 operational data calculation

NASA Astrophysics Data System (ADS)

Frybort, Jan

2017-09-01

Safe operation of a nuclear reactor requires an extensive calculational support. Operational data are determined by full-core calculations during the design phase of a fuel loading. Loading pattern and design of fuel assemblies are adjusted to meet safety requirements and optimize reactor operation. Nodal diffusion code ANDREA is used for this task in case of Czech VVER-1000 reactors. Nuclear data for this diffusion code are prepared regularly by lattice code HELIOS. These calculations are conducted in 2D on fuel assembly level. There is also possibility to calculate these macroscopic data by Monte-Carlo Serpent code. It can make use of alternative evaluated libraries. All calculations are affected by inherent uncertainties in nuclear data. It is useful to see results of full-core calculations based on two sets of diffusion data obtained by Serpent code calculations with ENDF/B-VII.1 and JEFF-3.2 nuclear data including also decay data library and fission yields data. The comparison is based directly on fuel assembly level macroscopic data and resulting operational data. This study illustrates effect of evaluated nuclear data library on full-core calculations of a large PWR reactor core. The level of difference which results exclusively from nuclear data selection can help to understand the level of inherent uncertainties of such full-core calculations.

MODFLOW 2.0: A program for predicting moderator flow patterns

NASA Astrophysics Data System (ADS)

Peterson, P. F.; Paik, I. K.

1991-07-01

Sudden changes in the temperature of flowing liquids can result in transient buoyancy forces which strongly impact the flow hydrodynamics via flow stratification. These effects have been studied for the case of potential flow of stratified liquids to line sinks, but not for moderator flow in SRS reactors. Standard codes, such as TRAC and COMMIX, do not have the capability to capture the stratification effect, due to strong numerical diffusion which smears away the hot/cold fluid interface. A related problem with standard codes is the inability to track plumes injected into the liquid flow, again due to numerical diffusion. The combined effects of buoyant stratification and plume dispersion have been identified as being important in the operation of the Supplementary Safety System which injects neutron-poison ink into SRS reactors to provide safe shutdown in the event of safety rod failure. The MODFLOW code discussed here provides transient moderator flow pattern information with stratification effects, and tracks the location of ink plumes in the reactor. The code, written in Fortran, is compiled for Macintosh II computers, and includes subroutines for interactive control and graphical output. Removing the graphics capabilities, the code can also be compiled on other computers. With graphics, in addition to the capability to perform safety related computations, MODFLOW also provides an easy tool for becoming familiar with flow distributions in SRS reactors.

Specifications for a coupled neutronics thermal-hydraulics SFR test case

NASA Astrophysics Data System (ADS)

Tassone, A.; Smirnov, A. D.; Tikhomirov, G. V.

2017-01-01

Coupling neutronics/thermal-hydraulics calculations for the design of nuclear reactors are a growing trend in the scientific community. This approach allows to properly represent the mutual feedbacks between the neutronic distribution and the thermal-hydraulics properties of the materials composing the reactor, details which are often lost when separate analysis are performed. In this work, a test case for a generation IV sodium-cooled fast reactor (SFR), based on the ASTRID concept developed by CEA, is proposed. Two sub-assemblies (SA) characterized by different fuel enrichment and layout are considered. Specifications for the test case are provided including geometrical data, material compositions, thermo-physical properties and coupling scheme details. Serpent and ANSYS-CFX are used as reference in the description of suitable inputs for the performing of the benchmark, but the use of other code combinations for the purpose of validation of the results is encouraged. The expected outcome of the test case are the axial distribution of volumetric power generation term (q‴), density and temperature for the fuel, the cladding and the coolant.

Mixed Legendre moments and discrete scattering cross sections for anisotropy representation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Calloo, A.; Vidal, J. F.; Le Tellier, R.

2012-07-01

This paper deals with the resolution of the integro-differential form of the Boltzmann transport equation for neutron transport in nuclear reactors. In multigroup theory, deterministic codes use transfer cross sections which are expanded on Legendre polynomials. This modelling leads to negative values of the transfer cross section for certain scattering angles, and hence, the multigroup scattering source term is wrongly computed. The first part compares the convergence of 'Legendre-expanded' cross sections with respect to the order used with the method of characteristics (MOC) for Pressurised Water Reactor (PWR) type cells. Furthermore, the cross section is developed using piecewise-constant functions, whichmore » better models the multigroup transfer cross section and prevents the occurrence of any negative value for it. The second part focuses on the method of solving the transport equation with the above-mentioned piecewise-constant cross sections for lattice calculations for PWR cells. This expansion thereby constitutes a 'reference' method to compare the conventional Legendre expansion to, and to determine its pertinence when applied to reactor physics calculations. (authors)« less

Procedure of recovery of pin-by-pin fields of energy release in the core of VVER-type reactor for the BIPR-8 code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gordienko, P. V., E-mail: gorpavel@vver.kiae.ru; Kotsarev, A. V.; Lizorkin, M. P.

2014-12-15

The procedure of recovery of pin-by-pin energy-release fields for the BIPR-8 code and the algorithm of the BIPR-8 code which is used in nodal computation of the reactor core and on which the recovery of pin-by-pin fields of energy release is based are briefly described. The description and results of the verification using the module of recovery of pin-by-pin energy-release fields and the TVS-M program are given.

150. ARAIII Reactor building (ARA608) Sections. Show highbay section, heater ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

150. ARA-III Reactor building (ARA-608) Sections. Show high-bay section, heater stack, and depth of reactor, piping, and heater pits. Aerojet-general 880-area/GCRE-608-A-3. Date: February 1958. Ineel index code no. 063-0608-00-013-102613. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Modelling the attenuation in the ATHENA finite elements code for the ultrasonic testing of austenitic stainless steel welds.

PubMed

Chassignole, B; Duwig, V; Ploix, M-A; Guy, P; El Guerjouma, R

2009-12-01

Multipass welds made in austenitic stainless steel, in the primary circuit of nuclear power plants with pressurized water reactors, are characterized by an anisotropic and heterogeneous structure that disturbs the ultrasonic propagation and makes ultrasonic non-destructive testing difficult. The ATHENA 2D finite element simulation code was developed to help understand the various physical phenomena at play. In this paper, we shall describe the attenuation model implemented in this code to give an account of wave scattering phenomenon through polycrystalline materials. This model is in particular based on the optimization of two tensors that characterize this material on the basis of experimental values of ultrasonic velocities attenuation coefficients. Three experimental configurations, two of which are representative of the industrial welds assessment case, are studied in view of validating the model through comparison with the simulation results. We shall thus provide a quantitative proof that taking into account the attenuation in the ATHENA code dramatically improves the results in terms of the amplitude of the echoes. The association of the code and detailed characterization of a weld's structure constitutes a remarkable breakthrough in the interpretation of the ultrasonic testing on this type of component.

Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baratta, A.J.

1997-07-01

To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts andmore » engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.« less

Coupled field effects in BWR stability simulations using SIMULATE-3K

DOE Office of Scientific and Technical Information (OSTI.GOV)

Borkowski, J.; Smith, K.; Hagrman, D.

1996-12-31

The SIMULATE-3K code is the transient analysis version of the Studsvik advanced nodal reactor analysis code, SIMULATE-3. Recent developments have focused on further broadening the range of transient applications by refinement of core thermal-hydraulic models and on comparison with boiling water reactor (BWR) stability measurements performed at Ringhals unit 1, during the startups of cycles 14 through 17.

Preliminary Analysis of the Transient Reactor Test Facility (TREAT) with PROTEUS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Connaway, H. M.; Lee, C. H.

The neutron transport code PROTEUS has been used to perform preliminary simulations of the Transient Reactor Test Facility (TREAT). TREAT is an experimental reactor designed for the testing of nuclear fuels and other materials under transient conditions. It operated from 1959 to 1994, when it was placed on non-operational standby. The restart of TREAT to support the U.S. Department of Energy’s resumption of transient testing is currently underway. Both single assembly and assembly-homogenized full core models have been evaluated. Simulations were performed using a historic set of WIMS-ANL-generated cross-sections as well as a new set of Serpent-generated cross-sections. To supportmore » this work, further analyses were also performed using additional codes in order to investigate particular aspects of TREAT modeling. DIF3D and the Monte-Carlo codes MCNP and Serpent were utilized in these studies. MCNP and Serpent were used to evaluate the effect of geometry homogenization on the simulation results and to support code-to-code comparisons. New meshes for the PROTEUS simulations were created using the CUBIT toolkit, with additional meshes generated via conversion of selected DIF3D models to support code-to-code verifications. All current analyses have focused on code-to-code verifications, with additional verification and validation studies planned. The analysis of TREAT with PROTEUS-SN is an ongoing project. This report documents the studies that have been performed thus far, and highlights key challenges to address in future work.« less

Core Dynamics Analysis for Reactivity Insertion and Loss of Coolant Flow Tests Using the High Temperature Engineering Test Reactor

NASA Astrophysics Data System (ADS)

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-temperature Gas-cooled Reactors (HTGRs). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named-ACCORD-, was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We have modified this code to use a model with four parallel channels and twenty temperature coefficients. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results. Moreover, the effect of the model is formulated quantitatively with our proposed equation. Finally, the pre-analytical result of the loss of coolant flow test by tripping all gas circulators is also discussed.

MTR plates modeling with MAIA

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marelle, V.; Dubois, S.; Ripert, M.

2008-07-15

MAIA is a thermo-mechanical code dedicated to the modeling of MTR fuel plates. The main physical phenomena modeled in the code are the cladding oxidation, the interaction between fuel and Al-matrix, the swelling due to fission products and the Al/fuel particles interaction. The creeping of the plate can be modeled in the mechanical calculation. MAIA has been validated on U-Mo dispersion fuel experiments such as IRIS 1 and 2 and FUTURE. The results are in rather good agreement with post-irradiation examinations. MAIA can also be used to calculate in-pile behavior of U{sub 3}Si{sub 2} plates as in the SHARE experimentmore » irradiated in the SCK/Mol BR2 reactor. The main outputs given by MAIA throughout the irradiation are temperatures, cladding oxidation thickness, interaction thickness, volume fraction of meat constituents, swelling, displacements, strains and stresses. MAIA is originally a two-dimensional code but a three-dimensional version is currently under development. (author)« less

139. ARAIII Index of drwaings of gascooled reactor experiment buildings. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

139. ARA-III Index of drwaings of gas-cooled reactor experiment buildings. Aerojet-general 880-area/GCRE-100. Date: February 1958. Ineel index code no. 063-9999-80-013-102505. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bily, T.

Thermoluminescent dosimeters represent very useful tool for gamma fields parameters measurements at nuclear research reactors, especially at zero power ones. {sup 7}LiF:Mg,Ti and {sup 7}LiF:Mg,Cu,P type TL dosimeters enable determination of only gamma component in mixed neutron - gamma field. At VR-1 reactor operated within the Faculty of Nuclear Sciences and Physical Engineering at the Czech Technical University in Prague the integral characteristics of gamma rays field were investigated, especially its spatial distribution and time behaviour, i.e. the non-saturated delayed gamma ray emission influence. Measured spatial distributions were compared with monte carlo code MCNP5 calculations. Although MCNP cannot generate delayedmore » gamma rays from fission, the relative gamma dose rate distribution is within {+-} 15% with measured values. The experiments were carried out with core configuration C1 consisting of LEU fuel IRT-4M (19.7 %). (author)« less

Spent fuel pool storage calculations using the ISOCRIT burnup credit tool

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kucukboyaci, Vefa; Marshall, William BJ J

2012-01-01

In order to conservatively apply burnup credit in spent fuel pool criticality safety analyses, Westinghouse has developed a software tool, ISOCRIT, for generating depletion isotopics. This tool is used to create isotopics data based on specific reactor input parameters, such as design basis assembly type; bounding power/burnup profiles; reactor specific moderator temperature profiles; pellet percent theoretical density; burnable absorbers, axial blanket regions, and bounding ppm boron concentration. ISOCRIT generates burnup dependent isotopics using PARAGON; Westinghouse's state-of-the-art and licensed lattice physics code. Generation of isotopics and passing the data to the subsequent 3D KENO calculations are performed in an automated fashion,more » thus reducing the chance for human error. Furthermore, ISOCRIT provides the means for responding to any customer request regarding re-analysis due to changed parameters (e.g., power uprate, exit temperature changes, etc.) with a quick turnaround.« less

Fundamentals, current state of the development of, and prospects for further improvement of the new-generation thermal-hydraulic computational HYDRA-IBRAE/LM code for simulation of fast reactor systems

NASA Astrophysics Data System (ADS)

Alipchenkov, V. M.; Anfimov, A. M.; Afremov, D. A.; Gorbunov, V. S.; Zeigarnik, Yu. A.; Kudryavtsev, A. V.; Osipov, S. L.; Mosunova, N. A.; Strizhov, V. F.; Usov, E. V.

2016-02-01

The conceptual fundamentals of the development of the new-generation system thermal-hydraulic computational HYDRA-IBRAE/LM code are presented. The code is intended to simulate the thermalhydraulic processes that take place in the loops and the heat-exchange equipment of liquid-metal cooled fast reactor systems under normal operation and anticipated operational occurrences and during accidents. The paper provides a brief overview of Russian and foreign system thermal-hydraulic codes for modeling liquid-metal coolants and gives grounds for the necessity of development of a new-generation HYDRA-IBRAE/LM code. Considering the specific engineering features of the nuclear power plants (NPPs) equipped with the BN-1200 and the BREST-OD-300 reactors, the processes and the phenomena are singled out that require a detailed analysis and development of the models to be correctly described by the system thermal-hydraulic code in question. Information on the functionality of the computational code is provided, viz., the thermalhydraulic two-phase model, the properties of the sodium and the lead coolants, the closing equations for simulation of the heat-mass exchange processes, the models to describe the processes that take place during the steam-generator tube rupture, etc. The article gives a brief overview of the usability of the computational code, including a description of the support documentation and the supply package, as well as possibilities of taking advantages of the modern computer technologies, such as parallel computations. The paper shows the current state of verification and validation of the computational code; it also presents information on the principles of constructing of and populating the verification matrices for the BREST-OD-300 and the BN-1200 reactor systems. The prospects are outlined for further development of the HYDRA-IBRAE/LM code, introduction of new models into it, and enhancement of its usability. It is shown that the program of development and practical application of the code will allow carrying out in the nearest future the computations to analyze the safety of potential NPP projects at a qualitatively higher level.

Analysis of supercritical CO{sub 2} cycle control strategies and dynamic response for Generation IV Reactors.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moisseytsev, A.; Sienicki, J. J.

2011-04-12

The analysis of specific control strategies and dynamic behavior of the supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle has been extended to the two reactor types selected for continued development under the Generation IV Nuclear Energy Systems Initiative; namely, the Very High Temperature Reactor (VHTR) and the Sodium-Cooled Fast Reactor (SFR). Direct application of the standard S-CO{sub 2} recompression cycle to the VHTR was found to be challenging because of the mismatch in the temperature drop of the He gaseous reactor coolant through the He-to-CO{sub 2} reactor heat exchanger (RHX) versus the temperature rise of the CO{sub 2} through themore » RHX. The reference VHTR features a large temperature drop of 450 C between the assumed core outlet and inlet temperatures of 850 and 400 C, respectively. This large temperature difference is an essential feature of the VHTR enabling a lower He flow rate reducing the required core velocities and pressure drop. In contrast, the standard recompression S-CO{sub 2} cycle wants to operate with a temperature rise through the RHX of about 150 C reflecting the temperature drop as the CO{sub 2} expands from 20 MPa to 7.4 MPa in the turbine and the fact that the cycle is highly recuperated such that the CO{sub 2} entering the RHX is effectively preheated. Because of this mismatch, direct application of the standard recompression cycle results in a relatively poor cycle efficiency of 44.9%. However, two approaches have been identified by which the S-CO{sub 2} cycle can be successfully adapted to the VHTR and the benefits of the S-CO{sub 2} cycle, especially a significant gain in cycle efficiency, can be realized. The first approach involves the use of three separate cascaded S-CO{sub 2} cycles. Each S-CO{sub 2} cycle is coupled to the VHTR through its own He-to-CO{sub 2} RHX in which the He temperature is reduced by 150 C. The three respective cycles have efficiencies of 54, 50, and 44%, respectively, resulting in a net cycle efficiency of 49.3 %. The other approach involves reducing the minimum cycle pressure significantly below the critical pressure such that the temperature drop in the turbine is increased while the minimum cycle temperature is maintained above the critical temperature to prevent the formation of a liquid phase. The latter approach also involves the addition of a precooler and a third compressor before the main compressor to retain the benefits of compression near the critical point with the main compressor. For a minimum cycle pressure of 1 MPa, a cycle efficiency of 49.5% is achieved. Either approach opens up the door to applying the SCO{sub 2} cycle to the VHTR. In contrast, the SFR system typically has a core outlet-inlet temperature difference of about 150 C such that the standard recompression cycle is ideally suited for direct application to the SFR. The ANL Plant Dynamics Code has been modified for application to the VHTR and SFR when the reactor side dynamic behavior is calculated with another system level computer code such as SAS4A/SYSSYS-1 in the SFR case. The key modification involves modeling heat exchange in the RHX, accepting time dependent tabular input from the reactor code, and generating time dependent tabular input to the reactor code such that both the reactor and S-CO{sub 2} cycle sides can be calculated in a convergent iterative scheme. This approach retains the modeling benefits provided by the detailed reactor system level code and can be applied to any reactor system type incorporating a S-CO{sub 2} cycle. This approach was applied to the particular calculation of a scram scenario for a SFR in which the main and intermediate sodium pumps are not tripped and the generator is not disconnected from the electrical grid in order to enhance heat removal from the reactor system thereby enhancing the cooldown rate of the Na-to-CO{sub 2} RHX. The reactor side is calculated with SAS4A/SASSYS-1 while the S-CO{sub 2} cycle is calculated with the Plant Dynamics Code with a number of iterations over a timescale of 500 seconds. It is found that the RHX undergoes a maximum cooldown rate of {approx} -0.3 C/s. The Plant Dynamics Code was also modified to decrease its running time by replacing the compressible flow form of the momentum equation with an incompressible flow equation for use inside of the cooler or recuperators where the CO{sub 2} has a compressibility similar to that of a liquid. Appendices provide a quasi-static control strategy for a SFR as well as the self-adaptive linear function fitting algorithm developed to produce the tabular data for input to the reactor code and Plant Dynamics Code from the detailed output of the other code.« less

148. ARAIII Reactor building (ARA608) Floor plan. Shows location of ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

148. ARA-III Reactor building (ARA-608) Floor plan. Shows location of reactor, heater, and mechanical loop pits; mechanical and electrical equipment rooms; and other work areas. Aerojet-general 880-area/GCRE-608-A-1. Date: February 1958. Ineel index code no. 063-0608-00-013-102612. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

158. ARAIII Reactor building (ARA608) Secondary cooling loop and piping ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

158. ARA-III Reactor building (ARA-608) Secondary cooling loop and piping plan. This drawing was selected as a typical example of piping arrangements within reactor building. Aerojet/general 880-area/GCRE-608-P-16. Date: February 1958. INeel index code no. 063-0608-50-013-102641. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Scoping Calculations of Power Sources for Nuclear Electric Propulsion

NASA Technical Reports Server (NTRS)

Difilippo, F. C.

1994-01-01

This technical memorandum describes models and calculational procedures to fully characterize the nuclear island of power sources for nuclear electric propulsion. Two computer codes were written: one for the gas-cooled NERVA derivative reactor and the other for liquid metal-cooled fuel pin reactors. These codes are going to be interfaced by NASA with the balance of plant in order to make scoping calculations for mission analysis.

Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

McDeavitt, Sean; Shao, Lin; Tsvetkov, Pavel

2014-04-07

Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. Many mechanistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, research, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development suchmore » that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.« less

MELCOR computer code manuals

DOE Office of Scientific and Technical Information (OSTI.GOV)

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

Five Lectures on Nuclear Reactors Presented at Cal Tech

DOE R&D Accomplishments Database

Weinberg, Alvin M.

1956-02-10

The basic issues involved in the physics and engineering of nuclear reactors are summarized. Topics discussed include theory of reactor design, technical problems in power reactors, physical problems in nuclear power production, and future developments in nuclear power. (C.H.)

OVERVIEW OF NUCLEAR PHYSICS LABORATORY (IMMEDIATELY EAST OF SPSE REACTOR ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

OVERVIEW OF NUCLEAR PHYSICS LABORATORY (IMMEDIATELY EAST OF SP-SE REACTOR ROOM), LEVEL -15’, LOOKING SOUTHWEST. NOTE SLIDING STEEL PLATE DOOR BETWEEN LABORATORY AND REACTOR ROOM - Physics Assembly Laboratory, Area A/M, Savannah River Site, Aiken, Aiken County, SC

Computation of Thermodynamic Equilibria Pertinent to Nuclear Materials in Multi-Physics Codes

NASA Astrophysics Data System (ADS)

Piro, Markus Hans Alexander

Nuclear energy plays a vital role in supporting electrical needs and fulfilling commitments to reduce greenhouse gas emissions. Research is a continuing necessity to improve the predictive capabilities of fuel behaviour in order to reduce costs and to meet increasingly stringent safety requirements by the regulator. Moreover, a renewed interest in nuclear energy has given rise to a "nuclear renaissance" and the necessity to design the next generation of reactors. In support of this goal, significant research efforts have been dedicated to the advancement of numerical modelling and computational tools in simulating various physical and chemical phenomena associated with nuclear fuel behaviour. This undertaking in effect is collecting the experience and observations of a past generation of nuclear engineers and scientists in a meaningful way for future design purposes. There is an increasing desire to integrate thermodynamic computations directly into multi-physics nuclear fuel performance and safety codes. A new equilibrium thermodynamic solver is being developed with this matter as a primary objective. This solver is intended to provide thermodynamic material properties and boundary conditions for continuum transport calculations. There are several concerns with the use of existing commercial thermodynamic codes: computational performance; limited capabilities in handling large multi-component systems of interest to the nuclear industry; convenient incorporation into other codes with quality assurance considerations; and, licensing entanglements associated with code distribution. The development of this software in this research is aimed at addressing all of these concerns. The approach taken in this work exploits fundamental principles of equilibrium thermodynamics to simplify the numerical optimization equations. In brief, the chemical potentials of all species and phases in the system are constrained by estimates of the chemical potentials of the system components at each iterative step, and the objective is to minimize the residuals of the mass balance equations. Several numerical advantages are achieved through this simplification. In particular, computational expense is reduced and the rate of convergence is enhanced. Furthermore, the software has demonstrated the ability to solve systems involving as many as 118 component elements. An early version of the code has already been integrated into the Advanced Multi-Physics (AMP) code under development by the Oak Ridge National Laboratory, Los Alamos National Laboratory, Idaho National Laboratory and Argonne National Laboratory. Keywords: Engineering, Nuclear -- 0552, Engineering, Material Science -- 0794, Chemistry, Mathematics -- 0405, Computer Science -- 0984

Qualification of Simulation Software for Safety Assessment of Sodium Cooled Fast Reactors. Requirements and Recommendations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Nicholas R.; Pointer, William David; Sieger, Matt

2016-04-01

The goal of this review is to enable application of codes or software packages for safety assessment of advanced sodium-cooled fast reactor (SFR) designs. To address near-term programmatic needs, the authors have focused on two objectives. First, the authors have focused on identification of requirements for software QA that must be satisfied to enable the application of software to future safety analyses. Second, the authors have collected best practices applied by other code development teams to minimize cost and time of initial code qualification activities and to recommend a path to the stated goal.

A physical description of fission product behavior fuels for advanced power reactors.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kaganas, G.; Rest, J.; Nuclear Engineering Division

2007-10-18

The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuelsmore » under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.« less

154. ARAIII Reactor building (ARA608) Foundation sections and details. Shows ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

154. ARA-III Reactor building (ARA-608) Foundation sections and details. Shows profiles of pits. Aerojet-general 888-area/GCRE-608-S-2. Date: February 1958. Ineel index code no. 062-0608-60-013-102654. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

153. ARAIII Reactor building (ARA608) Foundation plan. Aerojetgeneral 880area/GCRE608S1. Date: ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

153. ARA-III Reactor building (ARA-608) Foundation plan. Aerojet-general 880-area/GCRE-608-S-1. Date: February 1958. Ineel index code no. 063-0608-60-013-102653. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

152. ARAIII Reactor building (ARA608) Details of heater and piping ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

152. ARA-III Reactor building (ARA-608) Details of heater and piping pits, including instrumentation plan. Aerojet-general 880-area/GCRE-608-T-18. Date: November 1958. Ineel index code no. 063-0608-25-013-102677. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

149. ARAIII Reactor building (ARA608) Exterior elevations, showing north, south, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

149. ARA-III Reactor building (ARA-608) Exterior elevations, showing north, south, east, and west. Aerojet-general 880-area/GCRE-608-A-6. Date: February 1958. Ineel index code no. 063-0608-00-013-102615. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

FY2017 Updates to the SAS4A/SASSYS-1 Safety Analysis Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fanning, T. H.

The SAS4A/SASSYS-1 safety analysis software is used to perform deterministic analysis of anticipated events as well as design-basis and beyond-design-basis accidents for advanced fast reactors. It plays a central role in the analysis of U.S. DOE conceptual designs, proposed test and demonstration reactors, and in domestic and international collaborations. This report summarizes the code development activities that have taken place during FY2017. Extensions to the void and cladding reactivity feedback models have been implemented, and Control System capabilities have been improved through a new virtual data acquisition system for plant state variables and an additional Block Signal for a variablemore » lag compensator to represent reactivity feedback for novel shutdown devices. Current code development and maintenance needs are also summarized in three key areas: software quality assurance, modeling improvements, and maintenance of related tools. With ongoing support, SAS4A/SASSYS-1 can continue to fulfill its growing role in fast reactor safety analysis and help solidify DOE’s leadership role in fast reactor safety both domestically and in international collaborations.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dinh, Nam; Athe, Paridhi; Jones, Christopher

The Virtual Environment for Reactor Applications (VERA) code suite is assessed in terms of capability and credibility against the Consortium for Advanced Simulation of Light Water Reactors (CASL) Verification and Validation Plan (presented herein) in the context of three selected challenge problems: CRUD-Induced Power Shift (CIPS), Departure from Nucleate Boiling (DNB), and Pellet-Clad Interaction (PCI). Capability refers to evidence of required functionality for capturing phenomena of interest while capability refers to the evidence that provides confidence in the calculated results. For this assessment, each challenge problem defines a set of phenomenological requirements against which the VERA software is assessed. Thismore » approach, in turn, enables the focused assessment of only those capabilities relevant to the challenge problem. The evaluation of VERA against the challenge problem requirements represents a capability assessment. The mechanism for assessment is the Sandia-developed Predictive Capability Maturity Model (PCMM) that, for this assessment, evaluates VERA on 8 major criteria: (1) Representation and Geometric Fidelity, (2) Physics and Material Model Fidelity, (3) Software Quality Assurance and Engineering, (4) Code Verification, (5) Solution Verification, (6) Separate Effects Model Validation, (7) Integral Effects Model Validation, and (8) Uncertainty Quantification. For each attribute, a maturity score from zero to three is assigned in the context of each challenge problem. The evaluation of these eight elements constitutes the credibility assessment for VERA.« less

SCALE Code System 6.2.2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rearden, Bradley T.; Jessee, Matthew Anderson

The SCALE Code System is a widely used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including 3 deterministic and 3 Monte Carlomore » radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results. SCALE 6.2 represents one of the most comprehensive revisions in the history of SCALE, providing several new capabilities and significant improvements in many existing features.« less

Development of a model and computer code to describe solar grade silicon production processes

NASA Technical Reports Server (NTRS)

Srivastava, R.; Gould, R. K.

1979-01-01

Mathematical models, and computer codes based on these models were developed which allow prediction of the product distribution in chemical reactors in which gaseous silicon compounds are converted to condensed phase silicon. The reactors to be modeled are flow reactors in which silane or one of the halogenated silanes is thermally decomposed or reacted with an alkali metal, H2 or H atoms. Because the product of interest is particulate silicon, processes which must be modeled, in addition to mixing and reaction of gas-phase reactants, include the nucleation and growth of condensed Si via coagulation, condensation, and heterogeneous reaction.

EBR-II Reactor Physics Benchmark Evaluation Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pope, Chad L.; Lum, Edward S; Stewart, Ryan

This report provides a reactor physics benchmark evaluation with associated uncertainty quantification for the critical configuration of the April 1986 Experimental Breeder Reactor II Run 138B core configuration.

Review of CTF s Fuel Rod Modeling Using FRAPCON-4.0 s Centerline Temperature Predictions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Toptan, Aysenur; Salko, Robert K; Avramova, Maria

Coolant Boiling in Rod Arrays Two Fluid (COBRA-TF), or CTF1 [1], is a nuclear thermal hydraulic subchannel code used throughout academia and industry. CTF s fuel rod modeling is originally developed for VIPRE code [2]. Its methodology is based on GAPCON [3] and FRAP [4] fuel performance codes, and material properties are included from MATPRO handbook [5]. This work focuses on review of CTF s fuel rod modeling to address shortcomings in CTF s temperature predictions. CTF is compared to FRAPCON which is U.S. NRC s steady-state fuel performance code for light-water reactor fuel rods. FRAPCON calculates the changes inmore » fuel rod variables and temperatures including the eects of cladding hoop strain, cladding oxidation, hydriding, fuel irradiation swelling, densification, fission gas release and rod internal gas pressure. It uses fuel, clad and gap material properties from MATPRO. Additionally, it has its own models for fission gas release, cladding corrosion and cladding hydrogen pickup. It allows finite dierence or finite element approaches for its mechanical model. In this study, FRAPCON-4.0 [6] is used as a reference fuel performance code. In comparison, Halden Reactor Data for IFA432 Rod 1 and Rod 3. CTF simulations are performed in two ways; informing CTF with gap conductance value from FRAPCON, and using CTF s dynamic gap conductance model. First case is chosen to show temperature is predicted correctly with CTF s models for thermal and cladding conductivities once gap conductance is provided. Latter is to review CTF s dynamic gap conductance model. These Halden test cases are selected to be representative of cases with and without any physical contact between fuel-pellet and clad while reviewing functionality of CTF s dynamic gap conductance model. Improving the CTF s dynamic gap conductance model will allow prediction of fuel and cladding thermo-mechanical behavior under irradiation, and better temperature feedbacks from CTF in transient calculations.« less

Thermomechanical analysis of fast-burst reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miller, J.D.

1994-08-01

Fast-burst reactors are designed to provide intense, short-duration pulses of neutrons. The fission reaction also produces extreme time-dependent heating of the nuclear fuel. An existing transient-dynamic finite element code was modified specifically to compute the time-dependent stresses and displacements due to thermal shock loads of reactors. Thermomechanical analysis was then applied to determine structural feasibility of various concepts for an EDNA-type reactor and to optimize the mechanical design of the new SPR III-M reactor.

A flooding induced station blackout analysis for a pressurized water reactor using the RISMC toolkit

DOE PAGES

Mandelli, Diego; Prescott, Steven; Smith, Curtis; ...

2015-05-17

In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code calledmore » NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. The impact of power uprate is determined in terms of both core damage probability and safety margins.« less

PEBBLE: a two-dimensional steady-state pebble bed reactor thermal hydraulics code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vondy, D.R.

1981-09-01

This report documents the local implementation of the PEBBLE code to treat the two-dimensional steady-state pebble bed reactor thermal hydraulics problem. This code is implemented as a module of a computation system used for reactor core history calculations. Given power density data, the geometric description in (RZ), and basic heat removal conditions and thermal properties, the coolant properties, flow conditions, and temperature distributions in the pebble fuel elements are predicted. The calculation is oriented to the continuous fueling, steady state condition with consideration of the effect of the high energy neutron flux exposure and temperature history on the thermal conductivity.more » The coolant flow conditions are calculated for the same geometry as used in the neutronics calculation, power density and fluence data being used directly, and temperature results are made available for subsequent use.« less

New Reactor Physics Benchmark Data in the March 2012 Edition of the IRPhEP Handbook

DOE Office of Scientific and Technical Information (OSTI.GOV)

John D. Bess; J. Blair Briggs; Jim Gulliford

2012-11-01

The International Reactor Physics Experiment Evaluation Project (IRPhEP) was established to preserve integral reactor physics experimental data, including separate or special effects data for nuclear energy and technology applications. Numerous experiments that have been performed worldwide, represent a large investment of infrastructure, expertise, and cost, and are valuable resources of data for present and future research. These valuable assets provide the basis for recording, development, and validation of methods. If the experimental data are lost, the high cost to repeat many of these measurements may be prohibitive. The purpose of the IRPhEP is to provide an extensively peer-reviewed set ofmore » reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. Contributors from around the world collaborate in the evaluation and review of selected benchmark experiments for inclusion in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) [1]. Several new evaluations have been prepared for inclusion in the March 2012 edition of the IRPhEP Handbook.« less

Standardized verification of fuel cycle modeling

DOE PAGES

Feng, B.; Dixon, B.; Sunny, E.; ...

2016-04-05

A nuclear fuel cycle systems modeling and code-to-code comparison effort was coordinated across multiple national laboratories to verify the tools needed to perform fuel cycle analyses of the transition from a once-through nuclear fuel cycle to a sustainable potential future fuel cycle. For this verification study, a simplified example transition scenario was developed to serve as a test case for the four systems codes involved (DYMOND, VISION, ORION, and MARKAL), each used by a different laboratory participant. In addition, all participants produced spreadsheet solutions for the test case to check all the mass flows and reactor/facility profiles on a year-by-yearmore » basis throughout the simulation period. The test case specifications describe a transition from the current US fleet of light water reactors to a future fleet of sodium-cooled fast reactors that continuously recycle transuranic elements as fuel. After several initial coordinated modeling and calculation attempts, it was revealed that most of the differences in code results were not due to different code algorithms or calculation approaches, but due to different interpretations of the input specifications among the analysts. Therefore, the specifications for the test case itself were iteratively updated to remove ambiguity and to help calibrate interpretations. In addition, a few corrections and modifications were made to the codes as well, which led to excellent agreement between all codes and spreadsheets for this test case. Although no fuel cycle transition analysis codes matched the spreadsheet results exactly, all remaining differences in the results were due to fundamental differences in code structure and/or were thoroughly explained. As a result, the specifications and example results are provided so that they can be used to verify additional codes in the future for such fuel cycle transition scenarios.« less

Towards a Consolidated Approach for the Assessment of Evaluation Models of Nuclear Power Reactors

DOE PAGES

Epiney, A.; Canepa, S.; Zerkak, O.; ...

2016-11-02

The STARS project at the Paul Scherrer Institut (PSI) has adopted the TRACE thermal-hydraulic (T-H) code for best-estimate system transient simulations of the Swiss Light Water Reactors (LWRs). For analyses involving interactions between system and core, a coupling of TRACE with the SIMULATE-3K (S3K) LWR core simulator has also been developed. In this configuration, the TRACE code and associated nuclear power reactor simulation models play a central role to achieve a comprehensive safety analysis capability. Thus, efforts have now been undertaken to consolidate the validation strategy by implementing a more rigorous and structured assessment approach for TRACE applications involving eithermore » only system T-H evaluations or requiring interfaces to e.g. detailed core or fuel behavior models. The first part of this paper presents the preliminary concepts of this validation strategy. The principle is to systematically track the evolution of a given set of predicted physical Quantities of Interest (QoIs) over a multidimensional parametric space where each of the dimensions represent the evolution of specific analysis aspects, including e.g. code version, transient specific simulation methodology and model "nodalisation". If properly set up, such environment should provide code developers and code users with persistent (less affected by user effect) and quantified information (sensitivity of QoIs) on the applicability of a simulation scheme (codes, input models, methodology) for steady state and transient analysis of full LWR systems. Through this, for each given transient/accident, critical paths of the validation process can be identified that could then translate into defining reference schemes to be applied for downstream predictive simulations. In order to illustrate this approach, the second part of this paper presents a first application of this validation strategy to an inadvertent blowdown event that occurred in a Swiss BWR/6. The transient was initiated by the spurious actuation of the Automatic Depressurization System (ADS). The validation approach progresses through a number of dimensions here: First, the same BWR system simulation model is assessed for different versions of the TRACE code, up to the most recent one. The second dimension is the "nodalisation" dimension, where changes to the input model are assessed. The third dimension is the "methodology" dimension. In this case imposed power and an updated TRACE core model are investigated. For each step in each validation dimension, a common set of QoIs are investigated. For the steady-state results, these include fuel temperatures distributions. For the transient part of the present study, the evaluated QoIs include the system pressure evolution and water carry-over into the steam line.« less

Assessment of the TRACE Reactor Analysis Code Against Selected PANDA Transient Data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zavisca, M.; Ghaderi, M.; Khatib-Rahbar, M.

2006-07-01

The TRACE (TRAC/RELAP Advanced Computational Engine) code is an advanced, best-estimate thermal-hydraulic program intended to simulate the transient behavior of light-water reactor systems, using a two-fluid (steam and water, with non-condensable gas), seven-equation representation of the conservation equations and flow-regime dependent constitutive relations in a component-based model with one-, two-, or three-dimensional elements, as well as solid heat structures and logical elements for the control system. The U.S. Nuclear Regulatory Commission is currently supporting the development of the TRACE code and its assessment against a variety of experimental data pertinent to existing and evolutionary reactor designs. This paper presents themore » results of TRACE post-test prediction of P-series of experiments (i.e., tests comprising the ISP-42 blind and open phases) conducted at the PANDA large-scale test facility in 1990's. These results show reasonable agreement with the reported test results, indicating good performance of the code and relevant underlying thermal-hydraulic and heat transfer models. (authors)« less

Determination of the NPP Kr\\vsko spent fuel decay heat

NASA Astrophysics Data System (ADS)

Kromar, Marjan; Kurinčič, Bojan

2017-07-01

Nuclear fuel is designed to support fission process in a reactor core. Some of the isotopes, formed during the fission, decay and produce decay heat and radiation. Accurate knowledge of the nuclide inventory producing decay heat is important after reactor shut down, during the fuel storage and subsequent reprocessing or disposal. In this paper possibility to calculate the fuel isotopic composition and determination of the fuel decay heat with the Serpent code is investigated. Serpent is a well-known Monte Carlo code used primarily for the calculation of the neutron transport in a reactor. It has been validated for the burn-up calculations. In the calculation of the fuel decay heat different set of isotopes is important than in the neutron transport case. Comparison with the Origen code is performed to verify that the Serpent is taking into account all isotopes important to assess the fuel decay heat. After the code validation, a sensitivity study is carried out. Influence of several factors such as enrichment, fuel temperature, moderator temperature (density), soluble boron concentration, average power, burnable absorbers, and burnup is analyzed.

75 FR 42469 - Firstenergy Nuclear Operating Company; Request for Licensing Action

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-21

... nuclear plant in Ohio, preventing the reactor from restarting until such time that the NRC determines... Commission's regulations. The request has been referred to the Director of the Office of Nuclear Reactor... of Nuclear Reactor Regulation. [FR Doc. 2010-17834 Filed 7-20-10; 8:45 am] BILLING CODE 7590-01-P ...

Multiscale Simulations of ALD in Cross Flow Reactors

DOE PAGES

Yanguas-Gil, Angel; Libera, Joseph A.; Elam, Jeffrey W.

2014-08-13

In this study, we have developed a multiscale simulation code that allows us to study the impact of surface chemistry on the coating of large area substrates with high surface area/high aspect-ratio features. Our code, based on open-source libraries, takes advantage of the ALD surface chemistry to achieve an extremely efficient two-way coupling between reactor and feature length scales, and it can provide simulated quartz crystal microbalance and mass spectrometry data at any point of the reactor. By combining experimental surface characterization with simple analysis of growth profiles in a tubular cross flow reactor, we are able to extract amore » minimal set of reactions to effectively model the surface chemistry, including the presence of spurious CVD, to evaluate the impact of surface chemistry on the coating of large, high surface area substrates.« less

Critical experiments at Sandia National Laboratories : technical meeting on low-power critical facilities and small reactors.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harms, Gary A.; Ford, John T.; Barber, Allison Delo

2010-11-01

Sandia National Laboratories (SNL) has conducted radiation effects testing for the Department of Energy (DOE) and other contractors supporting the DOE since the 1960's. Over this period, the research reactor facilities at Sandia have had a primary mission to provide appropriate nuclear radiation environments for radiation testing and qualification of electronic components and other devices. The current generation of reactors includes the Annular Core Research Reactor (ACRR), a water-moderated pool-type reactor, fueled by elements constructed from UO2-BeO ceramic fuel pellets, and the Sandia Pulse Reactor III (SPR-III), a bare metal fast burst reactor utilizing a uranium-molybdenum alloy fuel. The SPR-IIImore » is currently defueled. The SPR Facility (SPRF) has hosted a series of critical experiments. A purpose-built critical experiment was first operated at the SPRF in the late 1980's. This experiment, called the Space Nuclear Thermal Propulsion Critical Experiment (CX), was designed to explore the reactor physics of a nuclear thermal rocket motor. This experiment was fueled with highly-enriched uranium carbide fuel in annular water-moderated fuel elements. The experiment program was completed and the fuel for the experiment was moved off-site. A second critical experiment, the Burnup Credit Critical Experiment (BUCCX) was operated at Sandia in 2002. The critical assembly for this experiment was based on the assembly used in the CX modified to accommodate low-enriched pin-type fuel in water moderator. This experiment was designed as a platform in which the reactivity effects of specific fission product poisons could be measured. Experiments were carried out on rhodium, an important fission product poison. The fuel and assembly hardware for the BUCCX remains at Sandia and is available for future experimentation. The critical experiment currently in operation at the SPRF is the Seven Percent Critical Experiment (7uPCX). This experiment is designed to provide benchmark reactor physics data to support validation of the reactor physics codes used to design commercial reactor fuel elements in an enrichment range above the current 5% enrichment cap. A first set of critical experiments in the 7uPCX has been completed. More experiments are planned in the 7uPCX series. The critical experiments at Sandia National Laboratories are currently funded by the US Department of Energy Nuclear Criticality Safety Program (NCSP). The NCSP has committed to maintain the critical experiment capability at Sandia and to support the development of a critical experiments training course at the facility. The training course is intended to provide hands-on experiment experience for the training of new and re-training of practicing Nuclear Criticality Safety Engineers. The current plans are for the development of the course to continue through the first part of fiscal year 2011 with the development culminating is the delivery of a prototype of the course in the latter part of the fiscal year. The course will be available in fiscal year 2012.« less

Hardware accelerated high performance neutron transport computation based on AGENT methodology

NASA Astrophysics Data System (ADS)

Xiao, Shanjie

The spatial heterogeneity of the next generation Gen-IV nuclear reactor core designs brings challenges to the neutron transport analysis. The Arbitrary Geometry Neutron Transport (AGENT) AGENT code is a three-dimensional neutron transport analysis code being developed at the Laboratory for Neutronics and Geometry Computation (NEGE) at Purdue University. It can accurately describe the spatial heterogeneity in a hierarchical structure through the R-function solid modeler. The previous version of AGENT coupled the 2D transport MOC solver and the 1D diffusion NEM solver to solve the three dimensional Boltzmann transport equation. In this research, the 2D/1D coupling methodology was expanded to couple two transport solvers, the radial 2D MOC solver and the axial 1D MOC solver, for better accuracy. The expansion was benchmarked with the widely applied C5G7 benchmark models and two fast breeder reactor models, and showed good agreement with the reference Monte Carlo results. In practice, the accurate neutron transport analysis for a full reactor core is still time-consuming and thus limits its application. Therefore, another content of my research is focused on designing a specific hardware based on the reconfigurable computing technique in order to accelerate AGENT computations. It is the first time that the application of this type is used to the reactor physics and neutron transport for reactor design. The most time consuming part of the AGENT algorithm was identified. Moreover, the architecture of the AGENT acceleration system was designed based on the analysis. Through the parallel computation on the specially designed, highly efficient architecture, the acceleration design on FPGA acquires high performance at the much lower working frequency than CPUs. The whole design simulations show that the acceleration design would be able to speedup large scale AGENT computations about 20 times. The high performance AGENT acceleration system will drastically shortening the computation time for 3D full-core neutron transport analysis, making the AGENT methodology unique and advantageous, and thus supplies the possibility to extend the application range of neutron transport analysis in either industry engineering or academic research.

Integration of Advanced Probabilistic Analysis Techniques with Multi-Physics Models

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cetiner, Mustafa Sacit; none,; Flanagan, George F.

2014-07-30

An integrated simulation platform that couples probabilistic analysis-based tools with model-based simulation tools can provide valuable insights for reactive and proactive responses to plant operating conditions. The objective of this work is to demonstrate the benefits of a partial implementation of the Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Detailed Framework Specification through the coupling of advanced PRA capabilities and accurate multi-physics plant models. Coupling a probabilistic model with a multi-physics model will aid in design, operations, and safety by providing a more accurate understanding of plant behavior. This represents the first attempt at actually integrating these two typesmore » of analyses for a control system used for operations, on a faster than real-time basis. This report documents the development of the basic communication capability to exchange data with the probabilistic model using Reliability Workbench (RWB) and the multi-physics model using Dymola. The communication pathways from injecting a fault (i.e., failing a component) to the probabilistic and multi-physics models were successfully completed. This first version was tested with prototypic models represented in both RWB and Modelica. First, a simple event tree/fault tree (ET/FT) model was created to develop the software code to implement the communication capabilities between the dynamic-link library (dll) and RWB. A program, written in C#, successfully communicates faults to the probabilistic model through the dll. A systems model of the Advanced Liquid-Metal Reactor–Power Reactor Inherently Safe Module (ALMR-PRISM) design developed under another DOE project was upgraded using Dymola to include proper interfaces to allow data exchange with the control application (ConApp). A program, written in C+, successfully communicates faults to the multi-physics model. The results of the example simulation were successfully plotted.« less

REACTOR PHYSICS QUARTERLY REPORT JANUARY, FEBRUARY, MARCH 1970

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schmid, L. C.; Clayton, E. D.; Heineman, R. E.

1970-05-01

The objective of the Reactor Physics Quarterly Report is to inform the scientific community in a timely manner of the technical progress made on the many phases of reactor physics work within the laboratory. The report contains brief technical discussions of accomplishments in all areas where significant progress has been made during the quarter.

Rigorous-two-Steps scheme of TRIPOLI-4® Monte Carlo code validation for shutdown dose rate calculation

NASA Astrophysics Data System (ADS)

Jaboulay, Jean-Charles; Brun, Emeric; Hugot, François-Xavier; Huynh, Tan-Dat; Malouch, Fadhel; Mancusi, Davide; Tsilanizara, Aime

2017-09-01

After fission or fusion reactor shutdown the activated structure emits decay photons. For maintenance operations the radiation dose map must be established in the reactor building. Several calculation schemes have been developed to calculate the shutdown dose rate. These schemes are widely developed in fusion application and more precisely for the ITER tokamak. This paper presents the rigorous-two-steps scheme implemented at CEA. It is based on the TRIPOLI-4® Monte Carlo code and the inventory code MENDEL. The ITER shutdown dose rate benchmark has been carried out, results are in a good agreement with the other participant.

Development of the V4.2m5 and V5.0m0 Multigroup Cross Section Libraries for MPACT for PWR and BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Kang Seog; Clarno, Kevin T.; Gentry, Cole

2017-03-01

The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CASL) core simulator is a 3-D whole core transport code being developed for the CASL toolset, Virtual Environment for Reactor Analysis (VERA). Key characteristics of the MPACT code include (1) a subgroup method for resonance selfshielding and (2) a whole-core transport solver with a 2-D/1-D synthesis method. The MPACT code requires a cross section library to support all the MPACT core simulation capabilities which would be the most influencing component for simulation accuracy.

Calculation and benchmarking of an azimuthal pressure vessel neutron fluence distribution using the BOXER code and scraping experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Holzgrewe, F.; Hegedues, F.; Paratte, J.M.

1995-03-01

The light water reactor BOXER code was used to determine the fast azimuthal neutron fluence distribution at the inner surface of the reactor pressure vessel after the tenth cycle of a pressurized water reactor (PWR). Using a cross-section library in 45 groups, fixed-source calculations in transport theory and x-y geometry were carried out to determine the fast azimuthal neutron flux distribution at the inner surface of the pressure vessel for four different cycles. From these results, the fast azimuthal neutron fluence after the tenth cycle was estimated and compared with the results obtained from scraping test experiments. In these experiments,more » small samples of material were taken from the inner surface of the pressure vessel. The fast neutron fluence was then determined form the measured activity of the samples. Comparing the BOXER and scraping test results have maximal differences of 15%, which is very good, considering the factor of 10{sup 3} neutron attenuation between the reactor core and the pressure vessel. To compare the BOXER results with an independent code, the 21st cycle of the PWR was also calculated with the TWODANT two-dimensional transport code, using the same group structure and cross-section library. Deviations in the fast azimuthal flux distribution were found to be <3%, which verifies the accuracy of the BOXER results.« less

MELCOR computer code manuals: Primer and user`s guides, Version 1.8.3 September 1994. Volume 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the US Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users` Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

Expert system for maintenance management of a boiling water reactor power plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hong Shen; Liou, L.W.; Levine, S.

1992-01-01

An expert system code has been developed for the maintenance of two boiling water reactor units in Berwick, Pennsylvania, that are operated by the Pennsylvania Power and Light Company (PP and L). The objective of this expert system code, where the knowledge of experienced operators and engineers is captured and implemented, is to support the decisions regarding which components can be safely and reliably removed from service for maintenance. It can also serve as a query-answering facility for checking the plant system status and for training purposes. The operating and maintenance information of a large number of support systems, whichmore » must be available for emergencies and/or in the event of an accident, is stored in the data base of the code. It identifies the relevant technical specifications and management rules for shutting down any one of the systems or removing a component from service to support maintenance. Because of the complexity and time needed to incorporate a large number of systems and their components, the first phase of the expert system develops a prototype code, which includes only the reactor core isolation coolant system, the high-pressure core injection system, the instrument air system, the service water system, and the plant electrical system. The next phase is scheduled to expand the code to include all other systems. This paper summarizes the prototype code and the design concept of the complete expert system code for maintenance management of all plant systems and components.« less

Critical Issues on Materials for Gen-IV Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Caro, M; Marian, J; Martinez, E

2009-02-27

Within the LDRD on 'Critical Issues on Materials for Gen-IV Reactors' basic thermodynamics of the Fe-Cr alloy and accurate atomistic modeling were used to help develop the capability to predict hardening, swelling and embrittlement using the paradigm of Multiscale Materials Modeling. Approaches at atomistic and mesoscale levels were linked to build-up the first steps in an integrated modeling platform that seeks to relate in a near-term effort dislocation dynamics to polycrystal plasticity. The requirements originated in the reactor systems under consideration today for future sources of nuclear energy. These requirements are beyond the present day performance of nuclear materials andmore » calls for the development of new, high temperature, radiation resistant materials. Fe-Cr alloys with 9-12% Cr content are the base matrix of advanced ferritic/martensitic (FM) steels envisaged as fuel cladding and structural components of Gen-IV reactors. Predictive tools are needed to calculate structural and mechanical properties of these steels. This project represents a contribution in that direction. The synergy between the continuous progress of parallel computing and the spectacular advances in the theoretical framework that describes materials have lead to a significant advance in our comprehension of materials properties and their mechanical behavior. We took this progress to our advantage and within this LDRD were able to provide a detailed physical understanding of iron-chromium alloys microstructural behavior. By combining ab-initio simulations, many-body interatomic potential development, and mesoscale dislocation dynamics we were able to describe their microstructure evolution. For the first time in the case of Fe-Cr alloys, atomistic and mesoscale were merged and the first steps taken towards incorporating ordering and precipitation effects into dislocation dynamics (DD) simulations. Molecular dynamics (MD) studies of the transport of self-interstitial, vacancy and point defect clusters in concentrated Fe-Cr alloys were performed for future diffusion data calculations. A recently developed parallel MC code with displacement allowed us to predict the evolution of the defect microstructures, local chemistry changes, grain boundary segregation and precipitation resulting from radiation enhanced diffusion. We showed that grain boundaries, dislocations and free surfaces are not preferential for alpha-prime precipitation, and explained experimental observations of short-range order (SRO) in Fe-rich FeCr alloys. Our atomistic studies of dislocation hardening allowed us to obtain dislocation mobility functions for BCC pure iron and Fe-Cr and determine for FCC metals the dislocation interaction with precipitates with a description to be used in Dislocation Dynamic (DD) codes. A Synchronous parallel Kinetic Monte Carlo code was developed and tested which promises to expand the range of applicability of kMC simulations. This LDRD furthered the limits of the available science on the thermodynamic and mechanic behavior of metallic alloys and extended the application of physically-based multiscale materials modeling to cases of severe temperature and neutron fluence conditions in advanced future nuclear reactors. The report is organized as follows: after a brief introduction, we present the research activities, and results obtained. We give recommendations on future LLNL activities that may contribute to the progress in this area, together with examples of possible research lines to be supported.« less

COMPLETE DETERMINATION OF POLARIZATION FOR A HIGH-ENERGY DEUTERON BEAM (thesis)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Button, J

1959-05-01

please delete the no. 17076<>13:017077The P/sub 1/ multigroup code was written for the IBM-704 in order to determine the accuracy of the few- group diffusion scheme with various imposed conditions and also to provide an alternate computational method when this scheme fails to be sufficiently accurate. The code solves for the spatially dependent multigroup flux, taking into account such nuclear phenomena is slowing down of neutrons resulting from elastic and inelastic scattering, the removal of neutrons resulting from epithermal capture and fission resonances, and the regeneration of fist neutrons resulting from fissioning which may occur in any of as manymore » as 80 fast multigroups or in the one thermal group. The code will accept as input a physical description of the reactor (that is: slab, cylindrical, or spherical geometry, number of points and regions, composition description group dependent boundary condition, transverse buckling, and mesh sizes) and a prepared library of nuclear properties of all the isotopes in each composition. The code will produce as output multigroup fluxes, currents, and isotopic slowing-down densities, in addition to pointwise and regionwise few-group macroscopic cross sections. (auth)« less

AREVA Developments for an Efficient and Reliable use of Monte Carlo codes for Radiation Transport Applications

NASA Astrophysics Data System (ADS)

Chapoutier, Nicolas; Mollier, François; Nolin, Guillaume; Culioli, Matthieu; Mace, Jean-Reynald

2017-09-01

In the context of the rising of Monte Carlo transport calculations for any kind of application, AREVA recently improved its suite of engineering tools in order to produce efficient Monte Carlo workflow. Monte Carlo codes, such as MCNP or TRIPOLI, are recognized as reference codes to deal with a large range of radiation transport problems. However the inherent drawbacks of theses codes - laboring input file creation and long computation time - contrast with the maturity of the treatment of the physical phenomena. The goals of the recent AREVA developments were to reach similar efficiency as other mature engineering sciences such as finite elements analyses (e.g. structural or fluid dynamics). Among the main objectives, the creation of a graphical user interface offering CAD tools for geometry creation and other graphical features dedicated to the radiation field (source definition, tally definition) has been reached. The computations times are drastically reduced compared to few years ago thanks to the use of massive parallel runs, and above all, the implementation of hybrid variance reduction technics. From now engineering teams are capable to deliver much more prompt support to any nuclear projects dealing with reactors or fuel cycle facilities from conceptual phase to decommissioning.

Methods and codes for neutronic calculations of the MARIA research reactor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andrzejewski, K.; Kulikowska, T.; Bretscher, M. M.

2002-02-18

The core of the MARIA high flux multipurpose research reactor is highly heterogeneous. It consists of beryllium blocks arranged in 6 x 8 matrix, tubular fuel assemblies, control rods and irradiation channels. The reflector is also heterogeneous and consists of graphite blocks clad with aluminum. Its structure is perturbed by the experimental beam tubes. This paper presents methods and codes used to calculate the MARIA reactor neutronics characteristics and experience gained thus far at IAE and ANL. At ANL the methods of MARIA calculations were developed in connection with the RERTR program. At IAE the package of programs was developedmore » to help its operator in optimization of fuel utilization.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Redi, M.H.; Mynick, H.E.; Suewattana, M.

Hamiltonian coordinate, guiding-center code calculations of the confinement of suprathermal ions in quasi-axisymmetric stellarator (QAS) designs have been carried out to evaluate the attractiveness of compact configurations which are optimized for ballooning stability. A new stellarator particle-following code is used to predict ion loss rates and particle confinement for thermal and neutral beam ions in a small experiment with R = 145 cm, B = 1-2 T and for alpha particles in a reactor-size device. In contrast to tokamaks, it is found that high edge poloidal flux has limited value in improving ion confinement in QAS, since collisional pitch-angle scatteringmore » drives ions into ripple wells and stochastic field regions, where they are quickly lost. The necessity for reduced stellarator ripple fields is emphasized. The high neutral beam ion loss predicted for these configurations suggests that more interesting physics could be explored with an experiment of less constrained size and magnetic field geometry.« less

Implementation, capabilities, and benchmarking of Shift, a massively parallel Monte Carlo radiation transport code

DOE PAGES

Pandya, Tara M.; Johnson, Seth R.; Evans, Thomas M.; ...

2015-12-21

This paper discusses the implementation, capabilities, and validation of Shift, a massively parallel Monte Carlo radiation transport package developed and maintained at Oak Ridge National Laboratory. It has been developed to scale well from laptop to small computing clusters to advanced supercomputers. Special features of Shift include hybrid capabilities for variance reduction such as CADIS and FW-CADIS, and advanced parallel decomposition and tally methods optimized for scalability on supercomputing architectures. Shift has been validated and verified against various reactor physics benchmarks and compares well to other state-of-the-art Monte Carlo radiation transport codes such as MCNP5, CE KENO-VI, and OpenMC. Somemore » specific benchmarks used for verification and validation include the CASL VERA criticality test suite and several Westinghouse AP1000 ® problems. These benchmark and scaling studies show promising results.« less

Versatile fusion source integrator AFSI for fast ion and neutron studies in fusion devices

NASA Astrophysics Data System (ADS)

Sirén, Paula; Varje, Jari; Äkäslompolo, Simppa; Asunta, Otto; Giroud, Carine; Kurki-Suonio, Taina; Weisen, Henri; JET Contributors, The

2018-01-01

ASCOT Fusion Source Integrator AFSI, an efficient tool for calculating fusion reaction rates and characterizing the fusion products, based on arbitrary reactant distributions, has been developed and is reported in this paper. Calculation of reactor-relevant D-D, D-T and D-3He fusion reactions has been implemented based on the Bosch-Hale fusion cross sections. The reactions can be calculated between arbitrary particle populations, including Maxwellian thermal particles and minority energetic particles. Reaction rate profiles, energy spectra and full 4D phase space distributions can be calculated for the non-isotropic reaction products. The code is especially suitable for integrated modelling in self-consistent plasma physics simulations as well as in the Serpent neutronics calculation chain. Validation of the model has been performed for neutron measurements at the JET tokamak and the code has been applied to predictive simulations in ITER.

Physical and chemical controls on the critical zone

USGS Publications Warehouse

Anderson, S.P.; Von Blanckenburg, F.; White, A.F.

2007-01-01

Geochemists have long recognized a correlation between rates of physical denudation and chemical weathering. What underlies this correlation? The Critical Zone can be considered as a feed-through reactor. Downward advance of the weathering front brings unweathered rock into the reactor. Fluids are supplied through precipitation. The reactor is stirred at the top by biological and physical processes. The balance between advance of the weathering front by mechanical and chemical processes and mass loss by denudation fixes the thickness of the Critical Zone reactor. The internal structure of this reactor is controlled by physical processes that create surface area, determine flow paths, and set the residence time of material in the Critical Zone. All of these impact chemical weathering flux.

Modeling and Analysis of Actinide Diffusion Behavior in Irradiated Metal Fuel

NASA Astrophysics Data System (ADS)

Edelmann, Paul G.

There have been numerous attempts to model fast reactor fuel behavior in the last 40 years. The US currently does not have a fully reliable tool to simulate the behavior of metal fuels in fast reactors. The experimental database necessary to validate the codes is also very limited. The DOE-sponsored Advanced Fuels Campaign (AFC) has performed various experiments that are ready for analysis. Current metal fuel performance codes are either not available to the AFC or have limitations and deficiencies in predicting AFC fuel performance. A modified version of a new fuel performance code, FEAST-Metal , was employed in this investigation with useful results. This work explores the modeling and analysis of AFC metallic fuels using FEAST-Metal, particularly in the area of constituent actinide diffusion behavior. The FEAST-Metal code calculations for this work were conducted at Los Alamos National Laboratory (LANL) in support of on-going activities related to sensitivity analysis of fuel performance codes. A sensitivity analysis of FEAST-Metal was completed to identify important macroscopic parameters of interest to modeling and simulation of metallic fuel performance. A modification was made to the FEAST-Metal constituent redistribution model to enable accommodation of newer AFC metal fuel compositions with verified results. Applicability of this modified model for sodium fast reactor metal fuel design is demonstrated.

REACTOR PHYSICS CONSTANTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1963-07-01

This second edition is based on data available on March 15, 1961. Sections on constants necessary for the interpretation of experimental data and on digital computer programs for reactor design and reactor physics have been added. 1344 references. (D.C.W.)

76 FR 35922 - Interim Staff Guidance Regarding the Environmental Report for Applications To Construct and/or...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-20

... Nuclear Reactor Regulation on the information that should be included in the Environmental Report, which...: Mr. Scott Sloan, Project Manager, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory..., Office of Nuclear Reactor Regulation. [FR Doc. 2011-15227 Filed 6-17-11; 8:45 am] BILLING CODE 7590-01-P ...

A parallel multi-domain solution methodology applied to nonlinear thermal transport problems in nuclear fuel pins

DOE PAGES

Philip, Bobby; Berrill, Mark A.; Allu, Srikanth; ...

2015-01-26

We describe an efficient and nonlinearly consistent parallel solution methodology for solving coupled nonlinear thermal transport problems that occur in nuclear reactor applications over hundreds of individual 3D physical subdomains. Efficiency is obtained by leveraging knowledge of the physical domains, the physics on individual domains, and the couplings between them for preconditioning within a Jacobian Free Newton Krylov method. Details of the computational infrastructure that enabled this work, namely the open source Advanced Multi-Physics (AMP) package developed by the authors are described. The details of verification and validation experiments, and parallel performance analysis in weak and strong scaling studies demonstratingmore » the achieved efficiency of the algorithm are presented. Moreover, numerical experiments demonstrate that the preconditioner developed is independent of the number of fuel subdomains in a fuel rod, which is particularly important when simulating different types of fuel rods. Finally, we demonstrate the power of the coupling methodology by considering problems with couplings between surface and volume physics and coupling of nonlinear thermal transport in fuel rods to an external radiation transport code.« less

Neutronics Analysis of SMART Small Modular Reactor using SRAC 2006 Code

NASA Astrophysics Data System (ADS)

Ramdhani, Rahmi N.; Prastyo, Puguh A.; Waris, Abdul; Widayani; Kurniadi, Rizal

2017-07-01

Small modular reactors (SMRs) are part of a new generation of nuclear reactor being developed worldwide. One of the advantages of SMR is the flexibility to adopt the advanced design concepts and technology. SMART (System integrated Modular Advanced ReacTor) is a small sized integral type PWR with a thermal power of 330 MW that has been developed by KAERI (Korea Atomic Energy Research Institute). SMART core consists of 57 fuel assemblies which are based on the well proven 17×17 array that has been used in Korean commercial PWRs. SMART is soluble boron free, and the high initial reactivity is mainly controlled by burnable absorbers. The goal of this study is to perform neutronics evaluation of SMART core with UO2 as main fuel. Neutronics calculation was performed by using PIJ and CITATION modules of SRAC 2006 code with JENDL 3.3 as nuclear data library.

FERRET-SAND II physics-dosimetry analysis for N Reactor Pressure Tubes 2954, 3053 and 1165 using a WIMS calculated input spectrum

DOE Office of Scientific and Technical Information (OSTI.GOV)

McElroy, W.N.; Kellogg, L.S.; Matsumoto, W.Y.

1988-05-01

This report is in response to a request from Westinghouse Hanford Company (WHC) that the PNL National Dosimetry Center (NDC) perform physics-dosimetry analyses (E > MeV) for N Reactor Pressure Tubes 2954 and 3053. As a result of these analyses, and recommendations for additional studies, two physics-dosimetry re-evaluations for Pressure Tube 1165 were also accomplished. The primary objective of Pacific Northwest Laboratories' (PNL) National Dosimetry Center (NDC) physics-dosimetry work for N Reactor was to provide FERRET-SAND II physics-dosimetry results to assist in the assessment of neutron radiation-induced changes in the physical and mechanical properties of N Reactor pressure tubes. 15more » refs., 6 figs., 5 tabs.« less

Assessment of MSIV full closure for Santa Maria de Garona Nuclear Power Plant using TRAC-BF1 (G1J1)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crespo, J.L.; Fernandez, R.A.

1993-06-01

This document presents a spurious Main Steam Isolation Value (MSIV) closure analysis for Santa Maria de Garorta Nuclear Power Plan describing the problems found when comparing calculated and real data. The plant is a General Electric Boiling Water Reactor 3, containment type Mark 1. It is operated by NUCLENOR, S.A. and was connected to the grid in 1971. The analysis has been performed by the Apphed Physics Department from the University of Cantabria and the Analysis and Operation Section from NUCLENOR, S.A. as a part of an agreement for developing an engineering simulator of operational transients and accidents for Santamore » Maria de Gamma Power Plant. The analysis was performed using the frozen version of TRAC-BFI (GlJl) code and is the second of two NUCLENOR contributions to the International Code Applications and Assessment Program (ICAP). The code was run in a Cyber 932 with operating system NOS/VE, property of NUCLENOR, S.A.. A programming effort was carried out in order to provide suitable graphics from the output file.« less

Rocketdyne/Westinghouse nuclear thermal rocket engine modeling

NASA Technical Reports Server (NTRS)

Glass, James F.

1993-01-01

The topics are presented in viewgraph form and include the following: systems approach needed for nuclear thermal rocket (NTR) design optimization; generic NTR engine power balance codes; rocketdyne nuclear thermal system code; software capabilities; steady state model; NTR engine optimizer code-logic; reactor power calculation logic; sample multi-component configuration; NTR design code output; generic NTR code at Rocketdyne; Rocketdyne NTR model; and nuclear thermal rocket modeling directions.

ON UPGRADING THE NUMERICS IN COMBUSTION CHEMISTRY CODES. (R824970)

EPA Science Inventory

A method of updating and reusing legacy FORTRAN codes for combustion simulations is presented using the DAEPACK software package. The procedure is demonstrated on two codes that come with the CHEMKIN-II package, CONP and SENKIN, for the constant-pressure batch reactor simulati...

Basic Nuclear Physics.

ERIC Educational Resources Information Center

Bureau of Naval Personnel, Washington, DC.

Basic concepts of nuclear structures, radiation, nuclear reactions, and health physics are presented in this text, prepared for naval officers. Applications to the area of nuclear power are described in connection with pressurized water reactors, experimental boiling water reactors, homogeneous reactor experiments, and experimental breeder…

SOPHAEROS code development and its application to falcon tests

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lajtha, G.; Missirlian, M.; Kissane, M.

1996-12-31

One of the key issues in source-term evaluation in nuclear reactor severe accidents is determination of the transport behavior of fission products released from the degrading core. The SOPHAEROS computer code is being developed to predict fission product transport in a mechanistic way in light water reactor circuits. These applications of the SOPHAEROS code to the Falcon experiments, among others not presented here, indicate that the numerical scheme of the code is robust, and no convergence problems are encountered. The calculation is also very fast being three times longer on a Sun SPARC 5 workstation than real time and typicallymore » {approx} 10 times faster than an identical calculation with the VICTORIA code. The study demonstrates that the SOPHAEROS 1.3 code is a suitable tool for prediction of the vapor chemistry and fission product transport with a reasonable level of accuracy. Furthermore, the fexibility of the code material data bank allows improvement of understanding of fission product transport and deposition in the circuit. Performing sensitivity studies with different chemical species or with different properties (saturation pressure, chemical equilibrium constants) is very straightforward.« less

Verification of combined thermal-hydraulic and heat conduction analysis code FLOWNET/TRUMP

NASA Astrophysics Data System (ADS)

Maruyama, Soh; Fujimoto, Nozomu; Kiso, Yoshihiro; Murakami, Tomoyuki; Sudo, Yukio

1988-09-01

This report presents the verification results of the combined thermal-hydraulic and heat conduction analysis code, FLOWNET/TRUMP which has been utilized for the core thermal hydraulic design, especially for the analysis of flow distribution among fuel block coolant channels, the determination of thermal boundary conditions for fuel block stress analysis and the estimation of fuel temperature in the case of fuel block coolant channel blockage accident in the design of the High Temperature Engineering Test Reactor(HTTR), which the Japan Atomic Energy Research Institute has been planning to construct in order to establish basic technologies for future advanced very high temperature gas-cooled reactors and to be served as an irradiation test reactor for promotion of innovative high temperature new frontier technologies. The verification of the code was done through the comparison between the analytical results and experimental results of the Helium Engineering Demonstration Loop Multi-channel Test Section(HENDEL T(sub 1-M)) with simulated fuel rods and fuel blocks.

Energy spectrum of 208Pb(n,x) reactions

NASA Astrophysics Data System (ADS)

Tel, E.; Kavun, Y.; Özdoǧan, H.; Kaplan, A.

2018-02-01

Fission and fusion reactor technologies have been investigated since 1950's on the world. For reactor technology, fission and fusion reaction investigations are play important role for improve new generation technologies. Especially, neutron reaction studies have an important place in the development of nuclear materials. So neutron effects on materials should study as theoretically and experimentally for improve reactor design. For this reason, Nuclear reaction codes are very useful tools when experimental data are unavailable. For such circumstances scientists created many nuclear reaction codes such as ALICE/ASH, CEM95, PCROSS, TALYS, GEANT, FLUKA. In this study we used ALICE/ASH, PCROSS and CEM95 codes for energy spectrum calculation of outgoing particles from Pb bombardment by neutron. While Weisskopf-Ewing model has been used for the equilibrium process in the calculations, full exciton, hybrid and geometry dependent hybrid nuclear reaction models have been used for the pre-equilibrium process. The calculated results have been discussed and compared with the experimental data taken from EXFOR.

Neutron dose rate analysis on HTGR-10 reactor using Monte Carlo code

NASA Astrophysics Data System (ADS)

Suwoto; Adrial, H.; Hamzah, A.; Zuhair; Bakhri, S.; Sunaryo, G. R.

2018-02-01

The HTGR-10 reactor is cylinder-shaped core fuelled with kernel TRISO coated fuel particles in the spherical pebble with helium cooling system. The outlet helium gas coolant temperature outputted from the reactor core is designed to 700 °C. One advantage HTGR type reactor is capable of co-generation, as an addition to generating electricity, the reactor was designed to produce heat at high temperature can be used for other processes. The spherical fuel pebble contains 8335 TRISO UO2 kernel coated particles with enrichment of 10% and 17% are dispersed in a graphite matrix. The main purpose of this study was to analysis the distribution of neutron dose rates generated from HTGR-10 reactors. The calculation and analysis result of neutron dose rate in the HTGR-10 reactor core was performed using Monte Carlo MCNP5v1.6 code. The problems of double heterogeneity in kernel fuel coated particles TRISO and spherical fuel pebble in the HTGR-10 core are modelled well with MCNP5v1.6 code. The neutron flux to dose conversion factors taken from the International Commission on Radiological Protection (ICRP-74) was used to determine the dose rate that passes through the active core, reflectors, core barrel, reactor pressure vessel (RPV) and a biological shield. The calculated results of neutron dose rate with MCNP5v1.6 code using a conversion factor of ICRP-74 (2009) for radiation workers in the radial direction on the outside of the RPV (radial position = 220 cm from the center of the patio HTGR-10) provides the respective value of 9.22E-4 μSv/h and 9.58E-4 μSv/h for enrichment 10% and 17%, respectively. The calculated values of neutron dose rates are compliant with BAPETEN Chairman’s Regulation Number 4 Year 2013 on Radiation Protection and Safety in Nuclear Energy Utilization which sets the limit value for the average effective dose for radiation workers 20 mSv/year or 10μSv/h. Thus the protection and safety for radiation workers to be safe from the radiation source has been fulfilled. From the result analysis, it can be concluded that the model of calculation result of neutron dose rate for HTGR-10 core has met the required radiation safety standards.

Handbook explaining the fundamentals of nuclear and atomic physics

NASA Technical Reports Server (NTRS)

Hanlen, D. F.; Morse, W. J.

1969-01-01

Indoctrination document presents nuclear, reactor, and atomic physics in an easy, straightforward manner. The entire subject of nuclear physics including atomic structure ionization, isotopes, radioactivity, and reactor dynamics is discussed.

A solid reactor core thermal model for nuclear thermal rockets

NASA Astrophysics Data System (ADS)

Rider, William J.; Cappiello, Michael W.; Liles, Dennis R.

1991-01-01

A Helium/Hydrogen Cooled Reactor Analysis (HERA) computer code has been developed. HERA has the ability to model arbitrary geometries in three dimensions, which allows the user to easily analyze reactor cores constructed of prismatic graphite elements. The code accounts for heat generation in the fuel, control rods, and other structures; conduction and radiation across gaps; convection to the coolant; and a variety of boundary conditions. The numerical solution scheme has been optimized for vector computers, making long transient analyses economical. Time integration is either explicit or implicit, which allows the use of the model to accurately calculate both short- or long-term transients with an efficient use of computer time. Both the basic spatial and temporal integration schemes have been benchmarked against analytical solutions.

The application of probabilistic fracture analysis to residual life evaluation of embrittled reactor vessels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dickson, T.L.; Simonen, F.A.

1992-05-01

Probabilistic fracture mechanics analysis is a major element of comprehensive probabilistic methodology on which current NRC regulatory requirements for pressurized water reactor vessel integrity evaluation are based. Computer codes such as OCA-P and VISA-II perform probabilistic fracture analyses to estimate the increase in vessel failure probability that occurs as the vessel material accumulates radiation damage over the operating life of the vessel. The results of such analyses, when compared with limits of acceptable failure probabilities, provide an estimation of the residual life of a vessel. Such codes can be applied to evaluate the potential benefits of plant-specific mitigating actions designedmore » to reduce the probability of failure of a reactor vessel. 10 refs.« less

The application of probabilistic fracture analysis to residual life evaluation of embrittled reactor vessels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dickson, T.L.; Simonen, F.A.

1992-01-01

Probabilistic fracture mechanics analysis is a major element of comprehensive probabilistic methodology on which current NRC regulatory requirements for pressurized water reactor vessel integrity evaluation are based. Computer codes such as OCA-P and VISA-II perform probabilistic fracture analyses to estimate the increase in vessel failure probability that occurs as the vessel material accumulates radiation damage over the operating life of the vessel. The results of such analyses, when compared with limits of acceptable failure probabilities, provide an estimation of the residual life of a vessel. Such codes can be applied to evaluate the potential benefits of plant-specific mitigating actions designedmore » to reduce the probability of failure of a reactor vessel. 10 refs.« less

EBR-II Static Neutronic Calculations by PHISICS / MCNP6 codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paolo Balestra; Carlo Parisi; Andrea Alfonsi

2016-02-01

The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) on the Shutdown Heat Removal Tests (SHRT) performed in the '80s at the Experimental fast Breeder Reactor EBR-II, USA. The scope of the CRP is to improve and validate the simulation tools for the study and the design of the liquid metal cooled fast reactors. Moreover, training of the next generation of fast reactor analysts is being also considered the other scope of the CRP. In this framework, a static neutronic model was developed, using state-of-the art neutron transport codes like SCALE/PHISICS (deterministic solution) and MCNP6 (stochastic solution).more » Comparison between both solutions is briefly illustrated in this summary.« less

Shielding calculation and criticality safety analysis of spent fuel transportation cask in research reactors.

PubMed

Mohammadi, A; Hassanzadeh, M; Gharib, M

2016-02-01

In this study, shielding calculation and criticality safety analysis were carried out for general material testing reactor (MTR) research reactors interim storage and relevant transportation cask. During these processes, three major terms were considered: source term, shielding, and criticality calculations. The Monte Carlo transport code MCNP5 was used for shielding calculation and criticality safety analysis and ORIGEN2.1 code for source term calculation. According to the results obtained, a cylindrical cask with body, top, and bottom thicknesses of 18, 13, and 13 cm, respectively, was accepted as the dual-purpose cask. Furthermore, it is shown that the total dose rates are below the normal transport criteria that meet the standards specified. Copyright © 2015 Elsevier Ltd. All rights reserved.

10 CFR Appendix D to Part 73 - Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Physical Protection of Irradiated Reactor Fuel in Transit... Irradiated Reactor Fuel in Transit, Training Program Subject Schedule Pursuant to the provision of § 73.37 of... reactor fuel is required to assure that individuals used as shipment escorts have completed a training...

10 CFR Appendix D to Part 73 - Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Physical Protection of Irradiated Reactor Fuel in Transit... Irradiated Reactor Fuel in Transit, Training Program Subject Schedule Pursuant to the provision of § 73.37 of... reactor fuel is required to assure that individuals used as shipment escorts have completed a training...

Fundamental approaches for analysis thermal hydraulic parameter for Puspati Research Reactor

NASA Astrophysics Data System (ADS)

Hashim, Zaredah; Lanyau, Tonny Anak; Farid, Mohamad Fairus Abdul; Kassim, Mohammad Suhaimi; Azhar, Noraishah Syahirah

2016-01-01

The 1-MW PUSPATI Research Reactor (RTP) is the one and only nuclear pool type research reactor developed by General Atomic (GA) in Malaysia. It was installed at Malaysian Nuclear Agency and has reached the first criticality on 8 June 1982. Based on the initial core which comprised of 80 standard TRIGA fuel elements, the very fundamental thermal hydraulic model was investigated during steady state operation using the PARET-code. The main objective of this paper is to determine the variation of temperature profiles and Departure of Nucleate Boiling Ratio (DNBR) of RTP at full power operation. The second objective is to confirm that the values obtained from PARET-code are in agreement with Safety Analysis Report (SAR) for RTP. The code was employed for the hot and average channels in the core in order to calculate of fuel's center and surface, cladding, coolant temperatures as well as DNBR's values. In this study, it was found that the results obtained from the PARET-code showed that the thermal hydraulic parameters related to safety for initial core which was cooled by natural convection was in agreement with the designed values and safety limit in SAR.

Comparison of computational results of the SABRE LMFBR pin bundle blockage code with data from well-instrumented out-of-pile test bundles (THORS bundles 3A and 5A)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dearing, J.F.

The Subchannel Analysis of Blockages in Reactor Elements (SABRE) computer code, developed by the United Kingdom Atomic Energy Authority, is currently the only practical tool available for performing detailed analyses of velocity and temperature fields in the recirculating flow regions downstream of blockages in liquid-metal fast breeder reactor (LMFBR) pin bundles. SABRE is a subchannel analysis code; that is, it accurately represents the complex geometry of nuclear fuel pins arranged on a triangular lattice. The results of SABRE computational models are compared here with temperature data from two out-of-pile 19-pin test bundles from the Thermal-Hydraulic Out-of-Reactor Safety (THORS) Facility atmore » Oak Ridge National Laboratory. One of these bundles has a small central flow blockage (bundle 3A), while the other has a large edge blockage (bundle 5A). Values that give best agreement with experiment for the empirical thermal mixing correlation factor, FMIX, in SABRE are suggested. These values of FMIX are Reynolds-number dependent, however, indicating that the coded turbulent mixing correlation is not appropriate for wire-wrap pin bundles.« less

THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vondy, D.R.

1984-07-01

The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures inmore » the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.« less

The Ongoing Impact of the U.S. Fast Reactor Integral Experiments Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

John D. Bess; Michael A. Pope; Harold F. McFarlane

2012-11-01

The creation of a large database of integral fast reactor physics experiments advanced nuclear science and technology in ways that were unachievable by less capital intensive and operationally challenging approaches. They enabled the compilation of integral physics benchmark data, validated (or not) analytical methods, and provided assurance of future rector designs The integral experiments performed at Argonne National Laboratory (ANL) represent decades of research performed to support fast reactor design and our understanding of neutronics behavior and reactor physics measurements. Experiments began in 1955 with the Zero Power Reactor No. 3 (ZPR-3) and terminated with the Zero Power Physics Reactormore » (ZPPR, originally the Zero Power Plutonium Reactor) in 1990 at the former ANL-West site in Idaho, which is now part of the Idaho National Laboratory (INL). Two additional critical assemblies, ZPR-6 and ZPR-9, operated at the ANL-East site in Illinois. A total of 128 fast reactor assemblies were constructed with these facilities [1]. The infrastructure and measurement capabilities are too expensive to be replicated in the modern era, making the integral database invaluable as the world pushes ahead with development of liquid metal cooled reactors.« less

Modeling of Flow Blockage in a Liquid Metal-Cooled Reactor Subassembly with a Subchannel Analysis Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jeong, Hae-Yong; Ha, Kwi-Seok; Chang, Won-Pyo

The local blockage in a subassembly of a liquid metal-cooled reactor (LMR) is of importance to the plant safety because of the compact design and the high power density of the core. To analyze the thermal-hydraulic parameters in a subassembly of a liquid metal-cooled reactor with a flow blockage, the Korea Atomic Energy Research Institute has developed the MATRA-LMR-FB code. This code uses the distributed resistance model to describe the sweeping flow formed by the wire wrap around the fuel rods and to model the recirculation flow after a blockage. The hybrid difference scheme is also adopted for the descriptionmore » of the convective terms in the recirculating wake region of low velocity. Some state-of-the-art turbulent mixing models were implemented in the code, and the models suggested by Rehme and by Zhukov are analyzed and found to be appropriate for the description of the flow blockage in an LMR subassembly. The MATRA-LMR-FB code predicts accurately the experimental data of the Oak Ridge National Laboratory 19-pin bundle with a blockage for both the high-flow and low-flow conditions. The influences of the distributed resistance model, the hybrid difference method, and the turbulent mixing models are evaluated step by step with the experimental data. The appropriateness of the models also has been evaluated through a comparison with the results from the COMMIX code calculation. The flow blockage for the KALIMER design has been analyzed with the MATRA-LMR-FB code and is compared with the SABRE code to guarantee the design safety for the flow blockage.« less

Deterministic Modeling of the High Temperature Test Reactor with DRAGON-HEXPEDITE

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. Ortensi; M.A. Pope; R.M. Ferrer

2010-10-01

The Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine the INL’s current prismatic reactor analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 fuel column thin annular core, and the fully loaded core critical condition with 30 fuel columns. Special emphasis is devoted to physical phenomena and artifacts in HTTR that are similar to phenomena and artifacts in themore » NGNP base design. The DRAGON code is used in this study since it offers significant ease and versatility in modeling prismatic designs. DRAGON can generate transport solutions via Collision Probability (CP), Method of Characteristics (MOC) and Discrete Ordinates (Sn). A fine group cross-section library based on the SHEM 281 energy structure is used in the DRAGON calculations. The results from this study show reasonable agreement in the calculation of the core multiplication factor with the MC methods, but a consistent bias of 2–3% with the experimental values is obtained. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement partially stems from the fact that during the experiments the control rods were adjusted to maintain criticality, whereas in the model, the rod positions were fixed. In addition, this work includes a brief study of a cross section generation approach that seeks to decouple the domain in order to account for neighbor effects. This spectral interpenetration is a dominant effect in annular HTR physics. This analysis methodology should be further explored in order to reduce the error that is systematically propagated in the traditional generation of cross sections.« less

Post-irradiation examination of uranium 7 wt% molybdenum atomized dispersion fuel

NASA Astrophysics Data System (ADS)

Leenaers, A.; Van den Berghe, S.; Koonen, E.; Jarousse, C.; Huet, F.; Trotabas, M.; Boyard, M.; Guillot, S.; Sannen, L.; Verwerft, M.

2004-10-01

Two low-enriched uranium fuel plates consisting of U-7wt%Mo atomized powder dispersed in an aluminum matrix, have been irradiated in the FUTURE irradiation rig of the BR2 reactor at SCK•CEN. The plates were submitted to a heat flux of maximum 353 W/cm 2 while the surface cladding temperature is kept below 130 °C. After 40 full power days, visual examination and profilometry of the fuel plates revealed an increase of the plate thickness. In view of this observation, the irradiation campaign was prematurely stopped and the fuel plates were retrieved from the reactor, having at their end-of-life a maximum burn-up of 32.8% 235U (6.5% FIMA). The microstructure of one of the fuel plates has been characterized in an extensive post-irradiation campaign. The U(Mo) fuel particles have been found to interact with the Al matrix, resulting in an interaction layer which can be identified as (U,Mo)Al 3 and (U,Mo)Al 4. Based on the composition of the interaction layer it is shown that the observed physical parameters like thickness of the interaction layer between the Al matrix and the U(Mo) fuel particles compare well to the values calculated by the MAIA code, an U(Mo) behavior modeling code developed by the Commissariat à l'énergie atomique (CEA).

Development of probabilistic design method for annular fuels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ozawa, Takayuki

2007-07-01

The increase of linear power and burn-up during the reactor operation is considered as one measure to ensure the utility of fast reactors in the future; for this the application of annular oxide fuels is under consideration. The annular fuel design code CEPTAR was developed in the Japan Atomic Energy Agency (JAEA) and verified by using many irradiation experiences with oxide fuels. In addition, the probabilistic fuel design code BORNFREE was also developed to provide a safe and reasonable fuel design and to evaluate the design margins quantitatively. This study aimed at the development of a probabilistic design method formore » annular oxide fuels; this was implemented in the developed BORNFREE-CEPTAR code, and the code was used to make a probabilistic evaluation with regard to the permissive linear power. (author)« less

PRIZMA predictions of in-core detection indications in the VVER-1000 reactor

NASA Astrophysics Data System (ADS)

Kandiev, Yadgar Z.; Kashayeva, Elena A.; Malyshin, Gennady N.; Modestov, Dmitry G.; Khatuntsev, Kirill E.

2014-06-01

The paper describes calculations which were done by the PRIZMA code(1) to predict indications of in-core rhodium detectors in the VVER-1000 reactor for some core fragments with allowance for fuel and rhodium burnout.

Multiphysics Object-Oriented Simulation Environment (MOOSE)

ScienceCinema

None

2017-12-09

Nuclear reactor operators can expand safety margins with more precise information about how materials behave inside operating reactors. INL's new simulation platform makes such studies easier & more informative by letting researchers "plug-n-play" their mathematical models, skipping years of computer code development.

Preliminary Study on LiF4-ThF4-PuF4 Utilization as Fuel Salt of miniFUJI Molten Salt Reactor

NASA Astrophysics Data System (ADS)

Waris, Abdul; Aji, Indarta K.; Pramuditya, Syeilendra; Widayani; Irwanto, Dwi

2016-08-01

miniFUJI reactor is molten salt reactor (MSR) which is one type of the Generation IV nuclear energy systems. The original miniFUJI reactor design uses LiF-BeF2-ThF4-233UF4 as a fuel salt. In the present study, the use of LiF4-ThF4-PuF4 as fuel salt instead of LiF-BeF2-ThF4-UF4 will be discussed. The neutronics cell calculation has been performed by using PIJ (collision probability method code) routine of SRAC 2006 code, with the nuclear data library is JENDL-4.0. The results reveal that the reactor can attain the criticality condition with the plutonium concentration in the fuel salt is equal to 9.16% or more. The conversion ratio diminishes with the enlarging of plutonium concentration in the fuel. The neutron spectrum of miniFUJI MSR with plutonium fuel becomes harder compared to that of the 233U fuel.

Effect of Light Water Reactor Water Environments on the Fatigue Life of Reactor Materials

DOE PAGES

Chopra, O. K.; Stevens, G. L.; Tregoning, R.; ...

2017-10-06

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) provides rules for the design of Class 1 components of nuclear power plants. Figures I-9.1 through I-9.6 of Appendix I to Section III of the Code specify fatigue design curves for applicable structural materials. However, the Code design curves do not explicitly address the effects of light water reactor (LWR) water environments. Existing fatigue strain-vs.-life (ε-N) laboratory data illustrate potentially significant effects of LWR water environments on the fatigue resistance of pressure vessel and piping steels. Extensive studies have been conducted at Argonne National Laboratory and elsewheremore » since 1990 to investigate the effects of LWR environments on the fatigue life of piping and pressure vessel steels. This article summarizes the results of these studies. Existing fatigue ε-N data were evaluated to identify the various material, environmental, and loading conditions that influence fatigue crack initiation; a methodology for estimating fatigue lives as a function of these parameters was developed. The effects were incorporated into the ASME Code Section III fatigue evaluations in terms of an environmental correction factor, F en, which is defined as the ratio of fatigue life in air at room temperature to the fatigue life in the LWR water environment at reactor operating temperatures. Available fatigue data were used to develop fatigue design curves for carbon and low-alloy steels, austenitic stainless steels, and nickel-chromium-iron (NiCr-Fe) alloys and their weld metals in air at room temperature. A review of the Code Section III fatigue adjustment factors of 2 on strain and 20 on life is also presented and the possible conservatism inherent in the choice of these adjustment factors is evaluated. A brief description of potential effects of neutron irradiation on fatigue crack initiation for these structural materials is also presented.« less

Effect of Light Water Reactor Water Environments on the Fatigue Life of Reactor Materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chopra, O. K.; Stevens, G. L.; Tregoning, R.

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) provides rules for the design of Class 1 components of nuclear power plants. Figures I-9.1 through I-9.6 of Appendix I to Section III of the Code specify fatigue design curves for applicable structural materials. However, the Code design curves do not explicitly address the effects of light water reactor (LWR) water environments. Existing fatigue strain-vs.-life (ε-N) laboratory data illustrate potentially significant effects of LWR water environments on the fatigue resistance of pressure vessel and piping steels. Extensive studies have been conducted at Argonne National Laboratory and elsewheremore » since 1990 to investigate the effects of LWR environments on the fatigue life of piping and pressure vessel steels. This article summarizes the results of these studies. Existing fatigue ε-N data were evaluated to identify the various material, environmental, and loading conditions that influence fatigue crack initiation; a methodology for estimating fatigue lives as a function of these parameters was developed. The effects were incorporated into the ASME Code Section III fatigue evaluations in terms of an environmental correction factor, F en, which is defined as the ratio of fatigue life in air at room temperature to the fatigue life in the LWR water environment at reactor operating temperatures. Available fatigue data were used to develop fatigue design curves for carbon and low-alloy steels, austenitic stainless steels, and nickel-chromium-iron (NiCr-Fe) alloys and their weld metals in air at room temperature. A review of the Code Section III fatigue adjustment factors of 2 on strain and 20 on life is also presented and the possible conservatism inherent in the choice of these adjustment factors is evaluated. A brief description of potential effects of neutron irradiation on fatigue crack initiation for these structural materials is also presented.« less

77 FR 30335 - Proposed Revision 3 to Standard Review Plan, Section 19.1 on Determining the Technical Adequacy...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-05-22

...). The Office of New Reactors and Office of Nuclear Reactor Regulation are revising SRP Section 19.1... of the Code of Federal Regulations (10 CFR), 50.71(h)(1), (h)(2), and (h)(3) for new reactors, (2... searching on http://www.regulations.gov under Docket ID NRC-2012-0113. You may submit comments by the...

CRAC2 model description

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ritchie, L.T.; Alpert, D.J.; Burke, R.P.

1984-03-01

The CRAC2 computer code is a revised version of CRAC (Calculation of Reactor Accident Consequences) which was developed for the Reactor Safety Study. This document provides an overview of the CRAC2 code and a description of each of the models used. Significant improvements incorporated into CRAC2 include an improved weather sequence sampling technique, a new evacuation model, and new output capabilities. In addition, refinements have been made to the atmospheric transport and deposition model. Details of the modeling differences between CRAC2 and CRAC are emphasized in the model descriptions.

Spheromak reactor-design study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Les, J.M.

1981-06-30

A general overview of spheromak reactor characteristics, such as MHD stability, start up, and plasma geometry is presented. In addition, comparisons are made between spheromaks, tokamaks and field reversed mirrors. The computer code Sphero is also discussed. Sphero is a zero dimensional time independent transport code that uses particle confinement times and profile parameters as input since they are not known with certainty at the present time. More specifically, Sphero numerically solves a given set of transport equations whose solutions include such variables as fuel ion (deuterium and tritium) density, electron density, alpha particle density and ion, electron temperatures.

Skyshine radiation from a pressurized water reactor containment dome

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peng, W.H.

1986-06-01

The radiation dose rates resulting from airborne activities inside a postaccident pressurized water reactor containment are calculated by a discrete ordinates/Monte Carlo combined method. The calculated total dose rates and the skyshine component are presented as a function of distance from the containment at three different elevations for various gamma-ray source energies. The one-dimensional (ANISN code) is used to approximate the skyshine dose rates from the hemisphere dome, and the results are compared favorably to more rigorous results calculated by a three-dimensional Monte Carlo code.

Reactivity Coefficient Calculation for AP1000 Reactor Using the NODAL3 Code

NASA Astrophysics Data System (ADS)

Pinem, Surian; Malem Sembiring, Tagor; Tukiran; Deswandri; Sunaryo, Geni Rina

2018-02-01

The reactivity coefficient is a very important parameter for inherent safety and stability of nuclear reactors operation. To provide the safety analysis of the reactor, the calculation of changes in reactivity caused by temperature is necessary because it is related to the reactor operation. In this paper, the temperature reactivity coefficients of fuel and moderator of the AP1000 core are calculated, as well as the moderator density and boron concentration. All of these coefficients are calculated at the hot full power condition (HFP). All neutron diffusion constant as a function of temperature, water density and boron concentration were generated by the SRAC2006 code. The core calculations for determination of the reactivity coefficient parameter are done by using NODAL3 code. The calculation results show that the fuel temperature, moderator temperature and boron reactivity coefficients are in the range between -2.613 pcm/°C to -4.657pcm/°C, -1.00518 pcm/°C to 1.00649 pcm/°C and -9.11361 pcm/ppm to -8.0751 pcm/ppm, respectively. For the water density reactivity coefficients, the positive reactivity occurs at the water temperature less than 190 °C. The calculation results show that the reactivity coefficients are accurate because the results have a very good agreement with the design value.

The MELTSPREAD Code for Modeling of Ex-Vessel Core Debris Spreading Behavior, Code Manual – Version3-beta

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farmer, M. T.

MELTSPREAD3 is a transient one-dimensional computer code that has been developed to predict the gravity-driven flow and freezing behavior of molten reactor core materials (corium) in containment geometries. Predictions can be made for corium flowing across surfaces under either dry or wet cavity conditions. The spreading surfaces that can be selected are steel, concrete, a user-specified material (e.g., a ceramic), or an arbitrary combination thereof. The corium can have a wide range of compositions of reactor core materials that includes distinct oxide phases (predominantly Zr, and steel oxides) plus metallic phases (predominantly Zr and steel). The code requires input thatmore » describes the containment geometry, melt “pour” conditions, and cavity atmospheric conditions (i.e., pressure, temperature, and cavity flooding information). For cases in which the cavity contains a preexisting water layer at the time of RPV failure, melt jet breakup and particle bed formation can be calculated mechanistically given the time-dependent melt pour conditions (input data) as well as the heatup and boiloff of water in the melt impingement zone (calculated). For core debris impacting either the containment floor or previously spread material, the code calculates the transient hydrodynamics and heat transfer which determine the spreading and freezing behavior of the melt. The code predicts conditions at the end of the spreading stage, including melt relocation distance, depth and material composition profiles, substrate ablation profile, and wall heatup. Code output can be used as input to other models such as CORQUENCH that evaluate long term core-concrete interaction behavior following the transient spreading stage. MELTSPREAD3 was originally developed to investigate BWR Mark I liner vulnerability, but has been substantially upgraded and applied to other reactor designs (e.g., the EPR), and more recently to the plant accidents at Fukushima Daiichi. The most recent round of improvements that are documented in this report have been specifically implemented to support industry in developing Severe Accident Water Management (SAWM) strategies for Boiling Water Reactors.« less

Identifying subassemblies by ultrasound to prevent fuel handling error in sodium fast reactors: First test performed in water

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paumel, Kevin; Lhuillier, Christian

2015-07-01

Identifying subassemblies by ultrasound is a method that is being considered to prevent handling errors in sodium fast reactors. It is based on the reading of a code (aligned notches) engraved on the subassembly head by an emitting/receiving ultrasonic sensor. This reading is carried out in sodium with high temperature transducers. The resulting one-dimensional C-scan can be likened to a binary code expressing the subassembly type and number. The first test performed in water investigated two parameters: width and depth of the notches. The code remained legible for notches as thin as 1.6 mm wide. The impact of the depthmore » seems minor in the range under investigation. (authors)« less

Methodes d'optimisation des parametres 2D du reflecteur dans un reacteur a eau pressurisee

NASA Astrophysics Data System (ADS)

Clerc, Thomas

With a third of the reactors in activity, the Pressurized Water Reactor (PWR) is today the most used reactor design in the world. This technology equips all the 19 EDF power plants. PWRs fit into the category of thermal reactors, because it is mainly the thermal neutrons that contribute to the fission reaction. The pressurized light water is both used as the moderator of the reaction and as the coolant. The active part of the core is composed of uranium, slightly enriched in uranium 235. The reflector is a region surrounding the active core, and containing mostly water and stainless steel. The purpose of the reflector is to protect the vessel from radiations, and also to slow down the neutrons and reflect them into the core. Given that the neutrons participate to the reaction of fission, the study of their behavior within the core is capital to understand the general functioning of how the reactor works. The neutrons behavior is ruled by the transport equation, which is very complex to solve numerically, and requires very long calculation. This is the reason why the core codes that will be used in this study solve simplified equations to approach the neutrons behavior in the core, in an acceptable calculation time. In particular, we will focus our study on the diffusion equation and approximated transport equations, such as SPN or S N equations. The physical properties of the reflector are radically different from those of the fissile core, and this structural change causes important tilt in the neutron flux at the core/reflector interface. This is why it is very important to accurately design the reflector, in order to precisely recover the neutrons behavior over the whole core. Existing reflector calculation techniques are based on the Lefebvre-Lebigot method. This method is only valid if the energy continuum of the neutrons is discretized in two energy groups, and if the diffusion equation is used. The method leads to the calculation of a homogeneous reflector. The aim of this study is to create a computational scheme able to compute the parameters of heterogeneous, multi-group reflectors, with both diffusion and SPN/SN operators. For this purpose, two computational schemes are designed to perform such a reflector calculation. The strategy used in both schemes is to minimize the discrepancies between a power distribution computed with a core code and a reference distribution, which will be obtained with an APOLLO2 calculation based on the method Method Of Characteristics (MOC). In both computational schemes, the optimization parameters, also called control variables, are the diffusion coefficients in each zone of the reflector, for diffusion calculations, and the P-1 corrected macroscopic total cross-sections in each zone of the reflector, for SPN/SN calculations (or correction factors on these parameters). After a first validation of our computational schemes, the results are computed, always by optimizing the fast diffusion coefficient for each zone of the reflector. All the tools of the data assimilation have been used to reflect the different behavior of the solvers in the different parts of the core. Moreover, the reflector is refined in six separated zones, corresponding to the physical structure of the reflector. There will be then six control variables for the optimization algorithms. [special characters omitted]. Our computational schemes are then able to compute heterogeneous, 2-group or multi-group reflectors, using diffusion or SPN/SN operators. The optimization performed reduces the discrepancies distribution between the power computed with the core codes and the reference power. However, there are two main limitations to this study: first the homogeneous modeling of the reflector assemblies doesn't allow to properly describe its physical structure near the core/reflector interface. Moreover, the fissile assemblies are modeled in infinite medium, and this model reaches its limit at the core/reflector interface. These two problems should be tackled in future studies. (Abstract shortened by UMI.).

SOC-DS computer code provides tool for design evaluation of homogeneous two-material nuclear shield

NASA Technical Reports Server (NTRS)

Disney, R. K.; Ricks, L. O.

1967-01-01

SOC-DS Code /Shield Optimization Code-Direc Search/, selects a nuclear shield material of optimum volume, weight, or cost to meet the requirments of a given radiation dose rate or energy transmission constraint. It is applicable to evaluating neutron and gamma ray shields for all nuclear reactors.

SASS-1--SUBASSEMBLY STRESS SURVEY CODE

DOE Office of Scientific and Technical Information (OSTI.GOV)

Friedrich, C.M.

1960-01-01

SASS-1, an IBM-704 FORTRAN code, calculates pressure, thermal, and combined stresses in a nuclear reactor core subassembly. In addition to cross- section stresses, the code calculates axial shear stresses needed to keep plane cross sections plane under axial variations of temperature. The input and output nomenclature, arrangement, and formats are described. (B.O.G.)

LSENS, The NASA Lewis Kinetics and Sensitivity Analysis Code

NASA Technical Reports Server (NTRS)

Radhakrishnan, K.

2000-01-01

A general chemical kinetics and sensitivity analysis code for complex, homogeneous, gas-phase reactions is described. The main features of the code, LSENS (the NASA Lewis kinetics and sensitivity analysis code), are its flexibility, efficiency and convenience in treating many different chemical reaction models. The models include: static system; steady, one-dimensional, inviscid flow; incident-shock initiated reaction in a shock tube; and a perfectly stirred reactor. In addition, equilibrium computations can be performed for several assigned states. An implicit numerical integration method (LSODE, the Livermore Solver for Ordinary Differential Equations), which works efficiently for the extremes of very fast and very slow reactions, is used to solve the "stiff" ordinary differential equation systems that arise in chemical kinetics. For static reactions, the code uses the decoupled direct method to calculate sensitivity coefficients of the dependent variables and their temporal derivatives with respect to the initial values of dependent variables and/or the rate coefficient parameters. Solution methods for the equilibrium and post-shock conditions and for perfectly stirred reactor problems are either adapted from or based on the procedures built into the NASA code CEA (Chemical Equilibrium and Applications).

The Modeling of Advanced BWR Fuel Designs with the NRC Fuel Depletion Codes PARCS/PATHS

DOE PAGES

Ward, Andrew; Downar, Thomas J.; Xu, Y.; ...

2015-04-22

The PATHS (PARCS Advanced Thermal Hydraulic Solver) code was developed at the University of Michigan in support of U.S. Nuclear Regulatory Commission research to solve the steady-state, two-phase, thermal-hydraulic equations for a boiling water reactor (BWR) and to provide thermal-hydraulic feedback for BWR depletion calculations with the neutronics code PARCS (Purdue Advanced Reactor Core Simulator). The simplified solution methodology, including a three-equation drift flux formulation and an optimized iteration scheme, yields very fast run times in comparison to conventional thermal-hydraulic systems codes used in the industry, while still retaining sufficient accuracy for applications such as BWR depletion calculations. Lastly, themore » capability to model advanced BWR fuel designs with part-length fuel rods and heterogeneous axial channel flow geometry has been implemented in PATHS, and the code has been validated against previously benchmarked advanced core simulators as well as BWR plant and experimental data. We describe the modifications to the codes and the results of the validation in this paper.« less

75 FR 67636 - Physical Protection of Shipments of Irradiated Reactor Fuel

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-03

...-2010-0340; Draft NUREG-0561, Revision 2] RIN 3150-AI64 Physical Protection of Shipments of Irradiated...- 0561, ``Physical Protection of Shipments of Irradiated Reactor Fuel.'' This document provides guidance to a licensee or applicant for implementation of proposed 10 CFR 73.37, ``Requirements for Physical...

78 FR 31821 - Physical Protection of Shipments of Irradiated Reactor Fuel

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-28

... NUCLEAR REGULATORY COMMISSION 10 CFR Part 73 [NRC-2010-0340; NRC-2009-0163] RIN 3150-AI64 Physical..., ``Physical Protection of Shipments of Irradiated Reactor Fuel.'' This revised document sets forth means... physical protection of spent nuclear fuel (SNF) during transportation by road, rail, and water; and for...

Evaluation of the finite element fuel rod analysis code (FRANCO)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, K.; Feltus, M.A.

1994-12-31

Knowledge of temperature distribution in a nuclear fuel rod is required to predict the behavior of fuel elements during operating conditions. The thermal and mechanical properties and performance characteristics are strongly dependent on the temperature, which can vary greatly inside the fuel rod. A detailed model of fuel rod behavior can be described by various numerical methods, including the finite element approach. The finite element method has been successfully used in many engineering applications, including nuclear piping and reactor component analysis. However, fuel pin analysis has traditionally been carried out with finite difference codes, with the exception of Electric Powermore » Research Institute`s FREY code, which was developed for mainframe execution. This report describes FRANCO, a finite element fuel rod analysis code capable of computing temperature disrtibution and mechanical deformation of a single light water reactor fuel rod.« less

Initial Comparison of Direct and Legacy Modeling Approaches for Radial Core Expansion Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shemon, Emily R.

2016-10-10

Radial core expansion in sodium-cooled fast reactors provides an important reactivity feedback effect. As the reactor power increases due to normal start up conditions or accident scenarios, the core and surrounding materials heat up, causing both grid plate expansion and bowing of the assembly ducts. When the core restraint system is designed correctly, the resulting structural deformations introduce negative reactivity which decreases the reactor power. Historically, an indirect procedure has been used to estimate the reactivity feedback due to structural deformation which relies upon perturbation theory and coupling legacy physics codes with limited geometry capabilities. With advancements in modeling andmore » simulation, radial core expansion phenomena can now be modeled directly, providing an assessment of the accuracy of the reactivity feedback coefficients generated by indirect legacy methods. Recently a new capability was added to the PROTEUS-SN unstructured geometry neutron transport solver to analyze deformed meshes quickly and directly. By supplying the deformed mesh in addition to the base configuration input files, PROTEUS-SN automatically processes material adjustments including calculation of region densities to conserve mass, calculation of isotopic densities according to material models (for example, sodium density as a function of temperature), and subsequent re-homogenization of materials. To verify the new capability of directly simulating deformed meshes, PROTEUS-SN was used to compute reactivity feedback for a series of contrived yet representative deformed configurations for the Advanced Burner Test Reactor design. The indirect legacy procedure was also performed to generate reactivity feedback coefficients for the same deformed configurations. Interestingly, the legacy procedure consistently overestimated reactivity feedbacks by 35% compared to direct simulations by PROTEUS-SN. This overestimation indicates that the legacy procedures are in fact not conservative and could be overestimating reactivity feedback effects that are closely tied to reactor safety. We conclude that there is indeed value in performing direct simulation of deformed meshes despite the increased computational expense. PROTEUS-SN is already part of the SHARP multi-physics toolkit where both thermal hydraulics and structural mechanical feedback modeling can be applied but this is the first comparison of direct simulation to legacy techniques for radial core expansion.« less

Development of the ANL plant dynamics code and control strategies for the supercritical carbon dioxide Brayton cycle and code validation with data from the Sandia small-scale supercritical carbon dioxide Brayton cycle test loop.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moisseytsev, A.; Sienicki, J. J.

2011-11-07

Significant progress has been made in the ongoing development of the Argonne National Laboratory (ANL) Plant Dynamics Code (PDC), the ongoing investigation and development of control strategies, and the analysis of system transient behavior for supercritical carbon dioxide (S-CO{sub 2}) Brayton cycles. Several code modifications have been introduced during FY2011 to extend the range of applicability of the PDC and to improve its calculational stability and speed. A new and innovative approach was developed to couple the Plant Dynamics Code for S-CO{sub 2} cycle calculations with SAS4A/SASSYS-1 Liquid Metal Reactor Code System calculations for the transient system level behavior onmore » the reactor side of a Sodium-Cooled Fast Reactor (SFR) or Lead-Cooled Fast Reactor (LFR). The new code system allows use of the full capabilities of both codes such that whole-plant transients can now be simulated without additional user interaction. Several other code modifications, including the introduction of compressor surge control, a new approach for determining the solution time step for efficient computational speed, an updated treatment of S-CO{sub 2} cycle flow mergers and splits, a modified enthalpy equation to improve the treatment of negative flow, and a revised solution of the reactor heat exchanger (RHX) equations coupling the S-CO{sub 2} cycle to the reactor, were introduced to the PDC in FY2011. All of these modifications have improved the code computational stability and computational speed, while not significantly affecting the results of transient calculations. The improved PDC was used to continue the investigation of S-CO{sub 2} cycle control and transient behavior. The coupled PDC-SAS4A/SASSYS-1 code capability was used to study the dynamic characteristics of a S-CO{sub 2} cycle coupled to a SFR plant. Cycle control was investigated in terms of the ability of the cycle to respond to a linear reduction in the electrical grid demand from 100% to 0% at a rate of 5%/minute. It was determined that utilization of turbine throttling control below 50% load improves the cycle efficiency significantly. Consequently, the cycle control strategy has been updated to include turbine throttle valve control. The new control strategy still relies on inventory control in the 50%-90% load range and turbine bypass for fine and fast generator output adjustments, but it now also includes turbine throttling control in the 0%-50% load range. In an attempt to investigate the feasibility of using the S-CO{sub 2} cycle for normal decay heat removal from the reactor, the cycle control study was extended beyond the investigation of normal load following. It was shown that such operation is possible with the extension of the inventory and the turbine throttling controls. However, the cycle operation in this range is calculated to be so inefficient that energy would need to be supplied from the electrical grid assuming that the generator could be capable of being operated in a motoring mode with an input electrical energy from the grid having a magnitude of about 20% of the nominal plant output electrical power level in order to maintain circulation of the CO{sub 2} in the cycle. The work on investigation of cycle operation at low power level will be continued in the future. In addition to the cycle control study, the coupled PDC-SAS4A/SASSYS-1 code system was also used to simulate thermal transients in the sodium-to-CO{sub 2} heat exchanger. Several possible conditions with the potential to introduce significant changes to the heat exchanger temperatures were identified and simulated. The conditions range from reactor scram and primary sodium pump failure or intermediate sodium pump failure on the reactor side to pipe breaks and valve malfunctions on the S-CO{sub 2} side. It was found that the maximum possible rate of the heat exchanger wall temperature change for the particular heat exchanger design assumed is limited to {+-}7 C/s for less than 10 seconds. Modeling in the Plant Dynamics Code has been compared with available data from the Sandia National Laboratories (SNL) small-scale S-CO{sub 2} Brayton cycle demonstration that is being assembled in a phased approach currently at Barber-Nichols Inc. and at SNL in the future. The available data was obtained with an earlier configuration of the S-CO{sub 2} loop involving only a single-turbo-alternator-compressor (TAC) instead of two TACs, a single low temperature recuperator (LTR) instead of both a LTR and a high temperature recuperator (HTR), and fewer than the later to be installed full set of electric heaters. Due to the absence of the full heating capability as well as the lack of a high temperature recuperator providing additional recuperation, the temperature conditions obtained with the loop are too low for the loop conditions to be prototypical of the S-CO{sub 2} cycle.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harmony, S.C.; Steiner, J.L.; Stumpf, H.J.

The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. As part of the preapplication and eventual design certification process, advanced reactor applicants are required to submit neutronic and thermal-hydraulic safety analyses over a sufficient range of normal operation, transient conditions, and specified accident sequences. Los Alamos is supporting the US Nuclear Regulatory Commission`s preapplication review of the PIUS reactor. A fully one-dimensional modelmore » of the PIUS reactor has been developed for the Transient Reactor Analysis Code, TRACPF1/MOD2. Early in 1992, ABB submitted a Supplemental Information Package describing recent design modifications. An important feature of the PIUS Supplement design was the addition of an active scram system that will function for most transient and accident conditions. A one-dimensional Transient Reactor Analysis Code baseline calculation of the PIUS Supplement design were performed for a break in the main steam line at the outlet nozzle of the loop 3 steam generator. Sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions following a main steam line break. The sensitivity study results provide insights into the robustness of the design.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, C.; Yu, G.; Wang, K.

The physical designs of the new concept reactors which have complex structure, various materials and neutronic energy spectrum, have greatly improved the requirements to the calculation methods and the corresponding computing hardware. Along with the widely used parallel algorithm, heterogeneous platforms architecture has been introduced into numerical computations in reactor physics. Because of the natural parallel characteristics, the CPU-FPGA architecture is often used to accelerate numerical computation. This paper studies the application and features of this kind of heterogeneous platforms used in numerical calculation of reactor physics through practical examples. After the designed neutron diffusion module based on CPU-FPGA architecturemore » achieves a 11.2 speed up factor, it is proved to be feasible to apply this kind of heterogeneous platform into reactor physics. (authors)« less

Eugene P. Wigner's Visionary Contributions to Generations-I through IV Fission Reactors

NASA Astrophysics Data System (ADS)

Carré, Frank

2014-09-01

Among Europe's greatest scientists who fled to Britain and America in the 1930s, Eugene P. Wigner made instrumental advances in reactor physics, reactor design and technology, and spent nuclear fuel processing for both purposes of developing atomic weapons during world-war II and nuclear power afterwards. Wigner who had training in chemical engineering and self-education in physics first gained recognition for his remarkable articles and books on applications of Group theory to Quantum mechanics, Solid state physics and other topics that opened new branches of Physics.

Low-power lead-cooled fast reactor loaded with MOX-fuel

NASA Astrophysics Data System (ADS)

Sitdikov, E. R.; Terekhova, A. M.

2017-01-01

Fast reactor for the purpose of implementation of research, education of undergraduate and doctoral students in handling innovative fast reactors and training specialists for atomic research centers and nuclear power plants (BRUTs) was considered. Hard neutron spectrum achieved in the fast reactor with compact core and lead coolant. Possibility of prompt neutron runaway of the reactor is excluded due to the low reactivity margin which is less than the effective fraction of delayed neutrons. The possibility of using MOX fuel in the BRUTs reactor was examined. The effect of Keff growth connected with replacement of natural lead coolant to 208Pb coolant was evaluated. The calculations and reactor core model were performed using the Serpent Monte Carlo code.

Extension of the TRANSURANUS burnup model to heavy water reactor conditions

NASA Astrophysics Data System (ADS)

Lassmann, K.; Walker, C. T.; van de Laar, J.

1998-06-01

The extension of the light water reactor burnup equations of the TRANSURANUS code to heavy water reactor conditions is described. Existing models for the fission of 235U and the buildup of plutonium in a heavy water reactor are evaluated. In order to overcome the limitations of the frequently used RADAR model at high burnup, a new model is presented. After verification against data for the radial distributions of Xe, Cs, Nd and Pu from electron probe microanalysis, the model is used to analyse the formation of the high burnup structure in a heavy water reactor. The new model allows the analysis of light water reactor fuel rod designs at high burnup in the OECD Halden Heavy Water Reactor.

LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carmack, William Jonathan; Barrett, Kristine Eloise; Chichester, Heather Jean MacLean

2015-09-01

The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirementsmore » for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.« less

Overview of Experiments for Physics of Fast Reactors from the International Handbooks of Evaluated Criticality Safety Benchmark Experiments and Evaluated Reactor Physics Benchmark Experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, J. D.; Briggs, J. B.; Gulliford, J.

Overview of Experiments to Study the Physics of Fast Reactors Represented in the International Directories of Critical and Reactor Experiments John D. Bess Idaho National Laboratory Jim Gulliford, Tatiana Ivanova Nuclear Energy Agency of the Organisation for Economic Cooperation and Development E.V.Rozhikhin, M.Yu.Sem?nov, A.M.Tsibulya Institute of Physics and Power Engineering The study the physics of fast reactors traditionally used the experiments presented in the manual labor of the Working Group on Evaluation of sections CSEWG (ENDF-202) issued by the Brookhaven National Laboratory in 1974. This handbook presents simplified homogeneous model experiments with relevant experimental data, as amended. The Nuclear Energymore » Agency of the Organization for Economic Cooperation and Development coordinates the activities of two international projects on the collection, evaluation and documentation of experimental data - the International Project on the assessment of critical experiments (1994) and the International Project on the assessment of reactor experiments (since 2005). The result of the activities of these projects are replenished every year, an international directory of critical (ICSBEP Handbook) and reactor (IRPhEP Handbook) experiments. The handbooks present detailed models of experiments with minimal amendments. Such models are of particular interest in terms of the settlements modern programs. The directories contain a large number of experiments which are suitable for the study of physics of fast reactors. Many of these experiments were performed at specialized critical stands, such as BFS (Russia), ZPR and ZPPR (USA), the ZEBRA (UK) and the experimental reactor JOYO (Japan), FFTF (USA). Other experiments, such as compact metal assembly, is also of interest in terms of the physics of fast reactors, they have been carried out on the universal critical stands in Russian institutes (VNIITF and VNIIEF) and the US (LANL, LLNL, and others.). Also worth mentioning is the critical experiments with fast reactor fuel rods in water, interesting in terms of justification of nuclear safety during transportation and storage of fresh and spent fuel. These reports provide a detailed review of the experiment, designate the area of their application and include results of calculations on modern systems of constants in comparison with the estimated experimental data.« less

Interface design of VSOP'94 computer code for safety analysis

NASA Astrophysics Data System (ADS)

Natsir, Khairina; Yazid, Putranto Ilham; Andiwijayakusuma, D.; Wahanani, Nursinta Adi

2014-09-01

Today, most software applications, also in the nuclear field, come with a graphical user interface. VSOP'94 (Very Superior Old Program), was designed to simplify the process of performing reactor simulation. VSOP is a integrated code system to simulate the life history of a nuclear reactor that is devoted in education and research. One advantage of VSOP program is its ability to calculate the neutron spectrum estimation, fuel cycle, 2-D diffusion, resonance integral, estimation of reactors fuel costs, and integrated thermal hydraulics. VSOP also can be used to comparative studies and simulation of reactor safety. However, existing VSOP is a conventional program, which was developed using Fortran 65 and have several problems in using it, for example, it is only operated on Dec Alpha mainframe platforms and provide text-based output, difficult to use, especially in data preparation and interpretation of results. We develop a GUI-VSOP, which is an interface program to facilitate the preparation of data, run the VSOP code and read the results in a more user friendly way and useable on the Personal 'Computer (PC). Modifications include the development of interfaces on preprocessing, processing and postprocessing. GUI-based interface for preprocessing aims to provide a convenience way in preparing data. Processing interface is intended to provide convenience in configuring input files and libraries and do compiling VSOP code. Postprocessing interface designed to visualized the VSOP output in table and graphic forms. GUI-VSOP expected to be useful to simplify and speed up the process and analysis of safety aspects.

Preliminary Analysis of the BASALA-H Experimental Programme

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blaise, Patrick; Fougeras, Philippe; Philibert, Herve

2002-07-01

This paper is focused on the preliminary analysis of results obtained on the first cores of the first phase of the BASALA (Boiling water reactor Advanced core physics Study Aimed at mox fuel Lattice) programme, aimed at studying the neutronic parameters in ABWR core in hot conditions, currently under investigation in the French EOLE critical facility, within the framework of a cooperation between NUPEC, CEA and Cogema. The first 'on-line' analysis of the results has been made, using a new preliminary design and safety scheme based on the French APOLLO-2 code in its 2.4 qualified version and associated CEA-93 V4more » (JEF-2.2) Library, that will enable the Experimental Physics Division (SPEx) to perform future core designs. It describes the scheme adopted and the results obtained in various cases, going to the critical size determination to the reactivity worth of the perturbed configurations (voided, over-moderated, and poisoned with Gd{sub 2}O{sub 3}-UO{sub 2} pins). A preliminary study on the experimental results on the MISTRAL-4 is resumed, and the comparison of APOLLO-2 versus MCNP-4C calculations on these cores is made. The results obtained show very good agreements between the two codes, and versus the experiment. This work opens the way to the future full analysis of the experimental results of the qualifying teams with completely validated schemes, based on the new 2.5 version of the APOLLO-2 code. (authors)« less

Neutrino Physics with Nuclear Reactors: An Overview

NASA Astrophysics Data System (ADS)

Ochoa-Ricoux, J. P.

Nuclear reactors provide an excellent environment for studying neutrinos and continue to play a critical role in unveiling the secrets of these elusive particles. A rich experimental program with reactor antineutrinos is currently ongoing, and leads the way in precision measurements of several oscillation parameters and in searching for new physics, such as the existence of light sterile neutrinos. Ongoing experiments have also been able to measure the flux and spectral shape of reactor antineutrinos with unprecedented statistics and as a function of core fuel evolution, uncovering anomalies that will lead to new physics and/or to an improved understanding of antineutrino emission from nuclear reactors. The future looks bright, with an aggressive program of next generation reactor neutrino experiments that will go after some of the biggest open questions in the field. This includes the JUNO experiment, the largest liquid scintillator detector ever constructed which will push the limits of this detection technology.

An approach to model reactor core nodalization for deterministic safety analysis

NASA Astrophysics Data System (ADS)

Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat @ Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd

2016-01-01

Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

An approach to model reactor core nodalization for deterministic safety analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my; Samsudin, Mohd Rafie, E-mail: rafies@tnb.com.my; Mamat Ibrahim, Mohd Rizal, E-mail: m-rizal@nuclearmalaysia.gov.my

Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to bemore » employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH{sub 1.6}, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D{sup ®} computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.« less

Education and training in the field of nuclear instrumentation and measurement: CEA/INSTN (National Institute for Nuclear Sciences and Technologies) strategy to improve and develop new pedagogical tools and methods

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vitart, Xavier; Foulon, Francois; Bodineau, Jean Christophe

Part of the French Alternative Energies and Atomic Energy Commission (CEA), the National Institute for Nuclear Science and Technology (INSTN) is a higher education institution whose mission is to provide students and professionals a high level of scientific and technological qualification in all disciplines related to nuclear energy applications. In this frame, INSTN carries out education and training (E and T) programs in nuclear instrumentation and radioprotection. Its strategy has always been to complete theoretical courses by training courses and laboratory works carried out on an extensive range of training tools that includes a large panel of nuclear instrumentation asmore » well as software applications. Since its creation in 1956, the INSTN has conducted both education and vocational programs on ionizing radiation detection. An extensive range of techniques have commonly been used during practical works with students and employees of companies who need to get the knowledge and specialization in this field. Today, the INSTN is mainly equipped with usual detectors and electronics in large numbers in order to be able to accommodate up to 48 trainees at the same time in two classrooms, with only two trainees for one workstation in order to optimize their learning. In the field of the neutron detection systems, the INSTN has strongly developed its offer taking advantage of the use of research reactors, such as ISIS reactor (700 kW) at Saclay. The implementation of neutron detection systems specific to the courses offers a unique way of observing and analysing the signal coming from neutron detectors, as well as learning how to set the parameters of the detection system in real conditions. Providing the trainees with an extensive overview of each part of the neutron monitoring instrumentation apply to a nuclear reactor, hands-on measurements on the ISIS reactor play a major role in ensuring a practical and comprehensive understanding of the neutron detection system and their integration in the safety system of nuclear reactors. It also gives a solid background for the follow up and the development of the neutron detection systems. Another field of activity of the INSTN is the development of new teaching tools using software applications. We describe hereafter two applications that have recently been developed by the institute, O.S.I.R.I.S and DOSIMEX. The INSTN and the company OREKA have developed an innovative teaching tool named O.S.I.R.I.S. The tool is built on a virtual 3D environment in which users operate in a totally free way. The action is located in a pressurised water reactor in the environment of a steam generator building. The users are students or professionals who want to learn and train on concrete situations on how to protect the workers against radiation. Through the use of this serious game, users must: establish a predictive dose evaluation, implement the principles of radiation protection, perform the dose result of the operation, analyse the gap between the predictive collective dose and the collective dose achieved, and decide about the lessons to provide feedback. Taking into account the increasing constraints related to radiation protection, it is essential to use computer codes to evaluate the dose rates generated by sources of ionizing radiation. Many powerful codes exist but are often relatively complex to implement and are dedicated to experts. In addition, these codes are often 'black boxes' that does not allow the user to understand the underlying physics. The objective of DOSIMEX codes developed by the INSTN is to give, by the means of a VBA code with a simple interface, the ability to easily conduct a wide range of radiological exposure situations: calculation of gamma, beta, neutron equivalent dose rate, calculation of internal contamination and atmospheric transfer.« less

75 FR 36126 - Office of New Reactors; Proposed Revision to Standard Review Plan Section 13.6.1, Revision 1 on...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-24

... NUCLEAR REGULATORY COMMISSION [NRC-2010-0228] Office of New Reactors; Proposed Revision to Standard Review Plan Section 13.6.1, Revision 1 on Physical Security--Combined License and Operating...), Section 13.6.1 on ``Physical Security--Combined License and Operating Reactors,'' (Agencywide Documents...

Numerical Tests for the Problem of U-Pu Fuel Burnup in Fuel Rod and Polycell Models Using the MCNP Code

NASA Astrophysics Data System (ADS)

Muratov, V. G.; Lopatkin, A. V.

An important aspect in the verification of the engineering techniques used in the safety analysis of MOX-fuelled reactors, is the preparation of test calculations to determine nuclide composition variations under irradiation and analysis of burnup problem errors resulting from various factors, such as, for instance, the effect of nuclear data uncertainties on nuclide concentration calculations. So far, no universally recognized tests have been devised. A calculation technique has been developed for solving the problem using the up-to-date calculation tools and the latest versions of nuclear libraries. Initially, in 1997, a code was drawn up in an effort under ISTC Project No. 116 to calculate the burnup in one VVER-1000 fuel rod, using the MCNP Code. Later on, the authors developed a computation technique which allows calculating fuel burnup in models of a fuel rod, or a fuel assembly, or the whole reactor. It became possible to apply it to fuel burnup in all types of nuclear reactors and subcritical blankets.

Results of comparative RBMK neutron computation using VNIIEF codes (cell computation, 3D statics, 3D kinetics). Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grebennikov, A.N.; Zhitnik, A.K.; Zvenigorodskaya, O.A.

1995-12-31

In conformity with the protocol of the Workshop under Contract {open_quotes}Assessment of RBMK reactor safety using modern Western Codes{close_quotes} VNIIEF performed a neutronics computation series to compare western and VNIIEF codes and assess whether VNIIEF codes are suitable for RBMK type reactor safety assessment computation. The work was carried out in close collaboration with M.I. Rozhdestvensky and L.M. Podlazov, NIKIET employees. The effort involved: (1) cell computations with the WIMS, EKRAN codes (improved modification of the LOMA code) and the S-90 code (VNIIEF Monte Carlo). Cell, polycell, burnup computation; (2) 3D computation of static states with the KORAT-3D and NEUmore » codes and comparison with results of computation with the NESTLE code (USA). The computations were performed in the geometry and using the neutron constants presented by the American party; (3) 3D computation of neutron kinetics with the KORAT-3D and NEU codes. These computations were performed in two formulations, both being developed in collaboration with NIKIET. Formulation of the first problem maximally possibly agrees with one of NESTLE problems and imitates gas bubble travel through a core. The second problem is a model of the RBMK as a whole with imitation of control and protection system controls (CPS) movement in a core.« less

76 FR 5102 - Draft NUREG-0561, Revision 2; Physical Protection of Shipments of Irradiated Reactor Fuel...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-28

... 3150-AI64 [NRC-2010-0340] Draft NUREG-0561, Revision 2; Physical Protection of Shipments of Irradiated...-0561, ``Physical Protection of Shipments of Irradiated Reactor Fuel.'' This document provides guidance on implementing the provisions of proposed 10 CFR Part 73.37, ``Requirements for Physical Protection...

Methodology, status and plans for development and assessment of the code ATHLET

DOE Office of Scientific and Technical Information (OSTI.GOV)

Teschendorff, V.; Austregesilo, H.; Lerchl, G.

1997-07-01

The thermal-hydraulic computer code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is being developed by the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) for the analysis of anticipated and abnormal plant transients, small and intermediate leaks as well as large breaks in light water reactors. The aim of the code development is to cover the whole spectrum of design basis and beyond design basis accidents (without core degradation) for PWRs and BWRs with only one code. The main code features are: advanced thermal-hydraulics; modular code architecture; separation between physical models and numerical methods; pre- and post-processing tools; portability. The codemore » has features that are of special interest for applications to small leaks and transients with accident management, e.g. initialization by a steady-state calculation, full-range drift-flux model, dynamic mixture level tracking. The General Control Simulation Module of ATHLET is a flexible tool for the simulation of the balance-of-plant and control systems including the various operator actions in the course of accident sequences with AM measures. The code development is accompained by a systematic and comprehensive validation program. A large number of integral experiments and separate effect tests, including the major International Standard Problems, have been calculated by GRS and by independent organizations. The ATHLET validation matrix is a well balanced set of integral and separate effects tests derived from the CSNI proposal emphasizing, however, the German combined ECC injection system which was investigated in the UPTF, PKL and LOBI test facilities.« less

Overview of the 2014 Edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook)

DOE Office of Scientific and Technical Information (OSTI.GOV)

John D. Bess; J. Blair Briggs; Jim Gulliford

2014-10-01

The International Reactor Physics Experiment Evaluation Project (IRPhEP) is a widely recognized world class program. The work of the IRPhEP is documented in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Integral data from the IRPhEP Handbook is used by reactor safety and design, nuclear data, criticality safety, and analytical methods development specialists, worldwide, to perform necessary validations of their calculational techniques. The IRPhEP Handbook is among the most frequently quoted reference in the nuclear industry and is expected to be a valuable resource for future decades.

Influence of Natural Convection and Thermal Radiation Multi-Component Transport in MOCVD Reactors

NASA Technical Reports Server (NTRS)

Lowry, S.; Krishnan, A.; Clark, I.

1999-01-01

The influence of Grashof and Reynolds number in Metal Organic Chemical Vapor (MOCVD) reactors is being investigated under a combined empirical/numerical study. As part of that research, the deposition of Indium Phosphide in an MOCVD reactor is modeled using the computational code CFD-ACE. The model includes the effects of convection, conduction, and radiation as well as multi-component diffusion and multi-step surface/gas phase chemistry. The results of the prediction are compared with experimental data for a commercial reactor and analyzed with respect to the model accuracy.

Prospective scenarios of nuclear energy evolution over the 21. century

DOE Office of Scientific and Technical Information (OSTI.GOV)

Massara, S.; Tetart, P.; Garzenne, C.

2006-07-01

In this paper, different world scenarios of nuclear energy development over the 21. century are analyzed, by means of the EDF fuel cycle simulation code for nuclear scenario studies, TIRELIRE - STRATEGIE. Three nuclear demand scenarios are considered, and the performance of different nuclear strategies in satisfying these scenarios is analyzed and discussed, focusing on natural uranium consumption and industrial requirements related to the nuclear reactors and the associated fuel cycle facilities. Both thermal-spectrum systems (Pressurized Water Reactor and High Temperature Gas-cooled Reactor) and Fast Reactors are investigated. (authors)

NEUTRON PHYSICS DIVISION ANNUAL PROGRESS REPORT. Period Ending September 1, 1962

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1963-01-11

A total of 74 subsections are included in the report. The information in 4 subsections was previously abstracted in NSA. Separate abstracts were prepared for 38 of the subsections. Those sections for which no abstracts were prepared contain information on prompt neutron lifetime, Rover critical experiments, Pu/sup 239/ fission, neutron decay, the O5R code, alpha scattering, 8 and P wavelengths, proton scattering, deuteron scattering, local optical potentials, N. S. Savamah radiation leakage, reactor shielding, cross section data analysis, gamma transport, gamma energy deposition, gaussian integration, data interpolation, neutron scattering, neutron energy deposition, space vehicles, computer analyses, shielding, positron sources, andmore » secondary particles. (J.R.D.)« less

A collision probability analysis of the double-heterogeneity problem

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hebert, A.

1993-10-01

A practical collision probability model is presented for the description of geometries with many levels of heterogeneity. Regular regions of the macrogeometry are assumed to contain a stochastic mixture of spherical grains or cylindrical tubes. Simple expressions for the collision probabilities in the global geometry are obtained as a function of the collision probabilities in the macro- and microgeometries. This model was successfully implemented in the collision probability kernel of the APOLLO-1, APOLLO-2, and DRAGON lattice codes for the description of a broad range of reactor physics problems. Resonance self-shielding and depletion calculations in the microgeometries are possible because eachmore » microregion is explicitly represented.« less

Numerical Simulation of the Emergency Condenser of the SWR-1000

DOE Office of Scientific and Technical Information (OSTI.GOV)

Krepper, Eckhard; Schaffrath, Andreas; Aszodi, Attila

The SWR-1000 is a new innovative boiling water reactor (BWR) concept, which was developed by Siemens AG. This concept is characterized in particular by passive safety systems (e.g., four emergency condensers, four building condensers, eight passive pressure pulse transmitters, and six gravity-driven core-flooding lines). In the framework of the BWR Physics and Thermohydraulic Complementary Action to the European Union BWR Research and Development Cluster, emergency condenser tests were performed by Forschungszentrum Juelich at the NOKO test facility. Posttest calculations with ATHLET are presented, which aim at the determination of the removable power of the emergency condenser and its operation mode.more » The one-dimensional thermal-hydraulic code ATHLET was extended by the module KONWAR for the calculation of the heat transfer coefficient during condensation in horizontal tubes. In addition, results of conventional finite difference calculations using the code CFX-4 are presented, which investigate the natural convection during the heatup process at the secondary side of the NOKO test facility.« less

Simulation of drift wave instability in field-reversed configurations using global magnetic geometry

NASA Astrophysics Data System (ADS)

Fulton, D. P.; Lau, C. K.; Lin, Z.; Tajima, T.; Holod, I.; the TAE Team

2016-10-01

Minimizing transport in the field-reversed configuration (FRC) is essential to enable FRC-based fusion reactors. Recently, significant progress on advanced beam-driven FRCs in C-2 and C-2U (at Tri Alpha Energy) provides opportunities to study transport properties using Doppler backscattering (DBS) measurements of turbulent fluctuations and kinetic particle-in-cell simulations of driftwaves in realistic equilibria via the Gyrokinetic Toroidal Code (GTC). Both measurements and simulations indicate relatively small fluctuations in the scrape-off layer (SOL). In the FRC core, local, single flux surface simulations reveal strong stabilization, while experiments indicate quiescent but finite fluctuations. One possible explanation is that turbulence may originate in the SOL and propagate at very low levels across the separatrix into the core. To test this hypothesis, a significant effort has been made to develop A New Code (ANC) based on GTC physics formulations, but using cylindrical coordinates which span the magnetic separatrix, including both core and SOL. Here, we present first results from global ANC simulations.

A DSMC Study of Low Pressure Argon Discharge

NASA Astrophysics Data System (ADS)

Hash, David; Meyyappan, M.

1997-10-01

Work toward a self-consistent plasma simulation using the DSMC method for examination of the flowfields of low-pressure high density plasma reactors is presented. Presently, DSMC simulations for these applications involve either treating the electrons as a fluid or imposing experimentally determined values for the electron number density profile. In either approach, the electrons themselves are not physically simulated. Self-consistent plasma DSMC simulations have been conducted for aerospace applications but at a severe computational cost due in part to the scalar architectures on which the codes were employed. The present work attempts to conduct such simulations at a more reasonable cost using a plasma version of the object-oriented parallel Cornell DSMC code, MONACO, on an IBM SP-2. Due the availability of experimental data, the GEC reference cell is chosen to conduct preliminary investigations. An argon discharge is examined thus affording a simple chemistry set with eight gas-phase reactions and five species: Ar, Ar^+, Ar^*, Ar_2, and e where Ar^* is a metastable.

Microdosimetric evaluation of the neutron field for BNCT at Kyoto University reactor by using the PHITS code.

PubMed

Baba, H; Onizuka, Y; Nakao, M; Fukahori, M; Sato, T; Sakurai, Y; Tanaka, H; Endo, S

2011-02-01

In this study, microdosimetric energy distributions of secondary charged particles from the (10)B(n,α)(7)Li reaction in boron-neutron capture therapy (BNCT) field were calculated using the Particle and Heavy Ion Transport code System (PHITS). The PHITS simulation was performed to reproduce the geometrical set-up of an experiment that measured the microdosimetric energy distributions at the Kyoto University Reactor where two types of tissue-equivalent proportional counters were used, one with A-150 wall alone and another with a 50-ppm-boron-loaded A-150 wall. It was found that the PHITS code is a useful tool for the simulation of the energy deposited in tissue in BNCT based on the comparisons with experimental results.

Verification of MCNP simulation of neutron flux parameters at TRIGA MK II reactor of Malaysia.

PubMed

Yavar, A R; Khalafi, H; Kasesaz, Y; Sarmani, S; Yahaya, R; Wood, A K; Khoo, K S

2012-10-01

A 3-D model for 1 MW TRIGA Mark II research reactor was simulated. Neutron flux parameters were calculated using MCNP-4C code and were compared with experimental results obtained by k(0)-INAA and absolute method. The average values of φ(th),φ(epi), and φ(fast) by MCNP code were (2.19±0.03)×10(12) cm(-2)s(-1), (1.26±0.02)×10(11) cm(-2)s(-1) and (3.33±0.02)×10(10) cm(-2)s(-1), respectively. These average values were consistent with the experimental results obtained by k(0)-INAA. The findings show a good agreement between MCNP code results and experimental results. Copyright © 2012 Elsevier Ltd. All rights reserved.

Impact of Reactor Operating Parameters on Cask Reactivity in BWR Burnup Credit

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ilas, Germina; Betzler, Benjamin R; Ade, Brian J

This paper discusses the effect of reactor operating parameters used in fuel depletion calculations on spent fuel cask reactivity, with relevance for boiling-water reactor (BWR) burnup credit (BUC) applications. Assessments that used generic BWR fuel assembly and spent fuel cask configurations are presented. The considered operating parameters, which were independently varied in the depletion simulations for the assembly, included fuel temperature, bypass water density, specific power, and operating history. Different operating history scenarios were considered for the assembly depletion to determine the effect of relative power distribution during the irradiation cycles, as well as the downtime between cycles. Depletion, decay,more » and criticality simulations were performed using computer codes and associated nuclear data within the SCALE code system. Results quantifying the dependence of cask reactivity on the assembly depletion parameters are presented herein.« less

Analysis of dosimetry from the H.B. Robinson unit 2 pressure vessel benchmark using RAPTOR-M3G and ALPAN

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fischer, G.A.

2011-07-01

Document available in abstract form only, full text of document follows: The dosimetry from the H. B. Robinson Unit 2 Pressure Vessel Benchmark is analyzed with a suite of Westinghouse-developed codes and data libraries. The radiation transport from the reactor core to the surveillance capsule and ex-vessel locations is performed by RAPTOR-M3G, a parallel deterministic radiation transport code that calculates high-resolution neutron flux information in three dimensions. The cross-section library used in this analysis is the ALPAN library, an Evaluated Nuclear Data File (ENDF)/B-VII.0-based library designed for reactor dosimetry and fluence analysis applications. Dosimetry is evaluated with the industry-standard SNLRMLmore » reactor dosimetry cross-section data library. (authors)« less

Experimental investigation of neutronic characteristics of the IR-8 reactor to confirm the results of calculations by MCU-PTR code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Surkov, A. V., E-mail: surkov.andrew@gmail.com; Kochkin, V. N.; Pesnya, Yu. E.

2015-12-15

A comparison of measured and calculated neutronic characteristics (fast neutron flux and fission rate of {sup 235}U) in the core and reflector of the IR-8 reactor is presented. The irradiation devices equipped with neutron activation detectors were prepared. The determination of fast neutron flux was performed using the {sup 54}Fe (n, p) and {sup 58}Ni (n, p) reactions. The {sup 235}U fission rate was measured using uranium dioxide with 10% enrichment in {sup 235}U. The determination of specific activities of detectors was carried out by measuring the intensity of characteristic gamma peaks using the ORTEC gamma spectrometer. Neutron fields inmore » the core and reflector of the IR-8 reactor were calculated using the MCU-PTR code.« less

Simulation of the hybrid and steady state advanced operating modes in ITER

NASA Astrophysics Data System (ADS)

Kessel, C. E.; Giruzzi, G.; Sips, A. C. C.; Budny, R. V.; Artaud, J. F.; Basiuk, V.; Imbeaux, F.; Joffrin, E.; Schneider, M.; Murakami, M.; Luce, T.; St. John, Holger; Oikawa, T.; Hayashi, N.; Takizuka, T.; Ozeki, T.; Na, Y.-S.; Park, J. M.; Garcia, J.; Tucillo, A. A.

2007-09-01

Integrated simulations are performed to establish a physics basis, in conjunction with present tokamak experiments, for the operating modes in the International Thermonuclear Experimental Reactor (ITER). Simulations of the hybrid mode are done using both fixed and free-boundary 1.5D transport evolution codes including CRONOS, ONETWO, TSC/TRANSP, TOPICS and ASTRA. The hybrid operating mode is simulated using the GLF23 and CDBM05 energy transport models. The injected powers are limited to the negative ion neutral beam, ion cyclotron and electron cyclotron heating systems. Several plasma parameters and source parameters are specified for the hybrid cases to provide a comparison of 1.5D core transport modelling assumptions, source physics modelling assumptions, as well as numerous peripheral physics modelling. Initial results indicate that very strict guidelines will need to be imposed on the application of GLF23, for example, to make useful comparisons. Some of the variations among the simulations are due to source models which vary widely among the codes used. In addition, there are a number of peripheral physics models that should be examined, some of which include fusion power production, bootstrap current, treatment of fast particles and treatment of impurities. The hybrid simulations project to fusion gains of 5.6-8.3, βN values of 2.1-2.6 and fusion powers ranging from 350 to 500 MW, under the assumptions outlined in section 3. Simulations of the steady state operating mode are done with the same 1.5D transport evolution codes cited above, except the ASTRA code. In these cases the energy transport model is more difficult to prescribe, so that energy confinement models will range from theory based to empirically based. The injected powers include the same sources as used for the hybrid with the possible addition of lower hybrid. The simulations of the steady state mode project to fusion gains of 3.5-7, βN values of 2.3-3.0 and fusion powers of 290 to 415 MW, under the assumptions described in section 4. These simulations will be presented and compared with particular focus on the resulting temperature profiles, source profiles and peripheral physics profiles. The steady state simulations are at an early stage and are focused on developing a range of safety factor profiles with 100% non-inductive current.

The TRIGA Reactor Facility at the Armed Forces Radiobiology Research Institute: A Simplified Technical Description.

DTIC Science & Technology

1986-05-01

COUNT Technical FROM_ TO May 1986 20 16. SUPPLEMENTARY NOTATION 17. COSATI CODES 18. SUBJECT TERMS iConitinue on reverse if neceasary and identify by...Reactor, Modes of Operation, The AFRRI Reactor, Exposure Facilities, and Cerenkov Radiation. I- 20 DISTRISUTIONIAVAILABILITY OF ABSTRACT 21. ABSTRACT...6 Exposure Facilities 12 Cerenkov Radiation 17 Acoessiofl For NTIS GRA&I DT.C TABUnamnnounced [] UusnriOfltond -. By IZ Distribution/ Availability

PBF Reactor Building (PER620) Cubicle 13. Plan, section, details. Note ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Reactor Building (PER-620) Cubicle 13. Plan, section, details. Note "quality assurance" code at bottom of drawing. Aerojet Nuclear Company. Date: May 1976. INEEL index no. 761-0620-00-400-195279 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

78 FR 9745 - Kewaunee Power Station; Application for Amendment to Facility Operating License

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-11

... FURTHER INFORMATION CONTACT: Karl Feintuch, Project Manager, Office of Nuclear Reactor Regulation, U.S... Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2013-03037 Filed 2-8-13; 8:45 am] BILLING CODE... NUCLEAR REGULATORY COMMISSION [Docket No. 50-305; NRC-2013-0028] Kewaunee Power Station...

Results of the Simulation of the HTR-Proteus Core 4.2 Using PEBBED-COMBINE: FY10 Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hans Gougar

2010-07-01

ABSTRACT The Idaho National Laboratory’s deterministic neutronics analysis codes and methods were applied to the computation of the core multiplication factor of the HTR-Proteus pebble bed reactor critical facility. This report is a follow-on to INL/EXT-09-16620 in which the same calculation was performed but using earlier versions of the codes and less developed methods. In that report, results indicated that the cross sections generated using COMBINE-7.0 did not yield satisfactory estimates of keff. It was concluded in the report that the modeling of control rods was not satisfactory. In the past year, improvements to the homogenization capability in COMBINE havemore » enabled the explicit modeling of TRIS particles, pebbles, and heterogeneous core zones including control rod regions using a new multi-scale version of COMBINE in which the 1-dimensional discrete ordinate transport code ANISN has been integrated. The new COMBINE is shown to yield benchmark quality results for pebble unit cell models, the first step in preparing few-group diffusion parameters for core simulations. In this report, the full critical core is modeled once again but with cross sections generated using the capabilities and physics of the improved COMBINE code. The new PEBBED-COMBINE model enables the exact modeling of the pebbles and control rod region along with better approximation to structures in the reflector. Initial results for the core multiplication factor indicate significant improvement in the INL’s tools for modeling the neutronic properties of a pebble bed reactor. Errors on the order of 1.6-2.5% in keff are obtained; a significant improvement over the 5-6% error observed in the earlier This is acceptable for a code system and model in the early stages of development but still too high for a production code. Analysis of a simpler core model indicates an over-prediction of the flux in the low end of the thermal spectrum. Causes of this discrepancy are under investigation. New homogenization techniques and assumptions were used in this analysis and as such, they require further confirmation and validation. Further refinement and review of the complex Proteus core model are likely to reduce the errors even further.« less

Non-Maxwellian fast particle effects in gyrokinetic GENE simulations

NASA Astrophysics Data System (ADS)

Di Siena, A.; Görler, T.; Doerk, H.; Bilato, R.; Citrin, J.; Johnson, T.; Schneider, M.; Poli, E.; JET Contributors

2018-04-01

Fast ions have recently been found to significantly impact and partially suppress plasma turbulence both in experimental and numerical studies in a number of scenarios. Understanding the underlying physics and identifying the range of their beneficial effect is an essential task for future fusion reactors, where highly energetic ions are generated through fusion reactions and external heating schemes. However, in many of the gyrokinetic codes fast ions are, for simplicity, treated as equivalent-Maxwellian-distributed particle species, although it is well known that to rigorously model highly non-thermalised particles, a non-Maxwellian background distribution function is needed. To study the impact of this assumption, the gyrokinetic code GENE has recently been extended to support arbitrary background distribution functions which might be either analytical, e.g., slowing down and bi-Maxwellian, or obtained from numerical fast ion models. A particular JET plasma with strong fast-ion related turbulence suppression is revised with these new code capabilities both with linear and nonlinear gyrokinetic simulations. It appears that the fast ion stabilization tends to be less strong but still substantial with more realistic distributions, and this improves the quantitative power balance agreement with experiments.

Nodal Green’s Function Method Singular Source Term and Burnable Poison Treatment in Hexagonal Geometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

A.A. Bingham; R.M. Ferrer; A.M. ougouag

2009-09-01

An accurate and computationally efficient two or three-dimensional neutron diffusion model will be necessary for the development, safety parameters computation, and fuel cycle analysis of a prismatic Very High Temperature Reactor (VHTR) design under Next Generation Nuclear Plant Project (NGNP). For this purpose, an analytical nodal Green’s function solution for the transverse integrated neutron diffusion equation is developed in two and three-dimensional hexagonal geometry. This scheme is incorporated into HEXPEDITE, a code first developed by Fitzpatrick and Ougouag. HEXPEDITE neglects non-physical discontinuity terms that arise in the transverse leakage due to the transverse integration procedure application to hexagonal geometry andmore » cannot account for the effects of burnable poisons across nodal boundaries. The test code being developed for this document accounts for these terms by maintaining an inventory of neutrons by using the nodal balance equation as a constraint of the neutron flux equation. The method developed in this report is intended to restore neutron conservation and increase the accuracy of the code by adding these terms to the transverse integrated flux solution and applying the nodal Green’s function solution to the resulting equation to derive a semi-analytical solution.« less

The virtual digital nuclear power plant: A modern tool for supporting the lifecycle of VVER-based nuclear power units

NASA Astrophysics Data System (ADS)

Arkadov, G. V.; Zhukavin, A. P.; Kroshilin, A. E.; Parshikov, I. A.; Solov'ev, S. L.; Shishov, A. V.

2014-10-01

The article describes the "Virtual Digital VVER-Based Nuclear Power Plant" computerized system comprising a totality of verified initial data (sets of input data for a model intended for describing the behavior of nuclear power plant (NPP) systems in design and emergency modes of their operation) and a unified system of new-generation computation codes intended for carrying out coordinated computation of the variety of physical processes in the reactor core and NPP equipment. Experiments with the demonstration version of the "Virtual Digital VVER-Based NPP" computerized system has shown that it is in principle possible to set up a unified system of computation codes in a common software environment for carrying out interconnected calculations of various physical phenomena at NPPs constructed according to the standard AES-2006 project. With the full-scale version of the "Virtual Digital VVER-Based NPP" computerized system put in operation, the concerned engineering, design, construction, and operating organizations will have access to all necessary information relating to the NPP power unit project throughout its entire lifecycle. The domestically developed commercial-grade software product set to operate as an independently operating application to the project will bring about additional competitive advantages in the modern market of nuclear power technologies.

VVER-440 and VVER-1000 reactor dosimetry benchmark - BUGLE-96 versus ALPAN VII.0

DOE Office of Scientific and Technical Information (OSTI.GOV)

Duo, J. I.

2011-07-01

Document available in abstract form only, full text of document follows: Analytical results of the vodo-vodyanoi energetichesky reactor-(VVER-) 440 and VVER-1000 reactor dosimetry benchmarks developed from engineering mockups at the Nuclear Research Inst. Rez LR-0 reactor are discussed. These benchmarks provide accurate determination of radiation field parameters in the vicinity and over the thickness of the reactor pressure vessel. Measurements are compared to calculated results with two sets of tools: TORT discrete ordinates code and BUGLE-96 cross-section library versus the newly Westinghouse-developed RAPTOR-M3G and ALPAN VII.0. The parallel code RAPTOR-M3G enables detailed neutron distributions in energy and space in reducedmore » computational time. ALPAN VII.0 cross-section library is based on ENDF/B-VII.0 and is designed for reactor dosimetry applications. It uses a unique broad group structure to enhance resolution in thermal-neutron-energy range compared to other analogous libraries. The comparison of fast neutron (E > 0.5 MeV) results shows good agreement (within 10%) between BUGLE-96 and ALPAN VII.O libraries. Furthermore, the results compare well with analogous results of participants of the REDOS program (2005). Finally, the analytical results for fast neutrons agree within 15% with the measurements, for most locations in all three mockups. In general, however, the analytical results underestimate the attenuation through the reactor pressure vessel thickness compared to the measurements. (authors)« less

Potential Effects of Leak-Before-Break on Light Water Reactor Design.

DTIC Science & Technology

1985-08-26

Boiler and Pressure Vessel Code . In fact, section 3 of that code was created for nuclear applications. This... Boiler and Pressure Vessel Code . The only major change which leak-before-break would require in these analyses would be that all piping to be considered...XI of the ASME Boiler and Pressure Vessel Code , and is already required for all Class I piping systems in the plant. Class I systems are those

Research and Engineering Operation, Irradiation Processing Department monthly record report, May 1965

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ambrose, T.W.

1965-06-04

Process and development activities reported include: depleted uranium irradiations, thoria irradiation, and hot die sizing. Reactor engineering activities include: brittle fracture of 190-C tanks, increased graphite temperature limits for the F reactor, VSR channel caulking, K reactor downcomer flow, zircaloy hydriding, and ribbed zircaloy process tubes. Reactor physics activities include: thoria irradiations, E-D irradiations, boiling protection with the high speed scanner, and in-core flux monitoring. Radiological engineering activities include: radiation control, classification, radiation occurrences, effluent activity data, and well car shielding. Process standards are listed, along with audits, and fuel failure experience. Operational physics and process physics studies are presented.more » Lastly, testing activities are detailed.« less

Interactive computer modeling of combustion chemistry and coalescence-dispersion modeling of turbulent combustion

NASA Technical Reports Server (NTRS)

Pratt, D. T.

1984-01-01

An interactive computer code for simulation of a high-intensity turbulent combustor as a single point inhomogeneous stirred reactor was developed from an existing batch processing computer code CDPSR. The interactive CDPSR code was used as a guide for interpretation and direction of DOE-sponsored companion experiments utilizing Xenon tracer with optical laser diagnostic techniques to experimentally determine the appropriate mixing frequency, and for validation of CDPSR as a mixing-chemistry model for a laboratory jet-stirred reactor. The coalescence-dispersion model for finite rate mixing was incorporated into an existing interactive code AVCO-MARK I, to enable simulation of a combustor as a modular array of stirred flow and plug flow elements, each having a prescribed finite mixing frequency, or axial distribution of mixing frequency, as appropriate. Further increase the speed and reliability of the batch kinetics integrator code CREKID was increased by rewriting in vectorized form for execution on a vector or parallel processor, and by incorporating numerical techniques which enhance execution speed by permitting specification of a very low accuracy tolerance.

Modified Laser and Thermos cell calculations on microcomputers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shapiro, A.; Huria, H.C.

1987-01-01

In the course of designing and operating nuclear reactors, many fuel pin cell calculations are required to obtain homogenized cell cross sections as a function of burnup. In the interest of convenience and cost, it would be very desirable to be able to make such calculations on microcomputers. In addition, such a microcomputer code would be very helpful for educational course work in reactor computations. To establish the feasibility of making detailed cell calculations on a microcomputer, a mainframe cell code was compiled and run on a microcomputer. The computer code Laser, originally written in Fortran IV for the IBM-7090more » class of mainframe computers, is a cylindrical, one-dimensional, multigroup lattice cell program that includes burnup. It is based on the MUFT code for epithermal and fast group calculations, and Thermos for the thermal calculations. There are 50 fast and epithermal groups and 35 thermal groups. Resonances are calculated assuming a homogeneous system and then corrected for self-shielding, Dancoff, and Doppler by self-shielding factors. The Laser code was converted to run on a microcomputer. In addition, the Thermos portion of Laser was extracted and compiled separately to have available a stand alone thermal code.« less

Thermal-mechanical performance modeling of thorium-plutonium oxide fuel and comparison with on-line irradiation data

NASA Astrophysics Data System (ADS)

Insulander Björk, Klara; Kekkonen, Laura

2015-12-01

Thorium-plutonium Mixed OXide (Th-MOX) fuel is considered for use in light water reactors fuel due to some inherent benefits over conventional fuel types in terms of neutronic properties. The good material properties of ThO2 also suggest benefits in terms of thermal-mechanical fuel performance, but the use of Th-MOX fuel for commercial power production demands that its thermal-mechanical behavior can be accurately predicted using a well validated fuel performance code. Given the scant operational experience with Th-MOX fuel, no such code is available today. This article describes the first phase of the development of such a code, based on the well-established code FRAPCON 3.4, and in particular the correlations reviewed and chosen for the fuel material properties. The results of fuel temperature calculations with the code in its current state of development are shown and compared with data from a Th-MOX test irradiation campaign which is underway in the Halden research reactor. The results are good for fresh fuel, whereas experimental complications make it difficult to judge the adequacy of the code for simulations of irradiated fuel.

10 CFR 73.35 - Requirements for physical protection of irradiated reactor fuel (100 grams or less) in transit.

Code of Federal Regulations, 2014 CFR

2014-01-01

... fuel (100 grams or less) in transit. 73.35 Section 73.35 Energy NUCLEAR REGULATORY COMMISSION... Transit § 73.35 Requirements for physical protection of irradiated reactor fuel (100 grams or less) in... quantity of irradiated reactor fuel weighing 100 grams (0.22 pounds) or less in net weight of irradiated...

Development and Assessment of CTF for Pin-resolved BWR Modeling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salko, Robert K; Wysocki, Aaron J; Collins, Benjamin S

2017-01-01

CTF is the modernized and improved version of the subchannel code, COBRA-TF. It has been adopted by the Consortium for Advanced Simulation for Light Water Reactors (CASL) for subchannel analysis applications and thermal hydraulic feedback calculations in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS). CTF is now jointly developed by Oak Ridge National Laboratory and North Carolina State University. Until now, CTF has been used for pressurized water reactor modeling and simulation in CASL, but in the future it will be extended to boiling water reactor designs. This required development activities to integrate the code into the VERA-CSmore » workflow and to make it more ecient for full-core, pin resolved simulations. Additionally, there is a significant emphasis on producing high quality tools that follow a regimented software quality assurance plan in CASL. Part of this plan involves performing validation and verification assessments on the code that are easily repeatable and tied to specific code versions. This work has resulted in the CTF validation and verification matrix being expanded to include several two-phase flow experiments, including the General Electric 3 3 facility and the BWR Full-Size Fine Mesh Bundle Tests (BFBT). Comparisons with both experimental databases is reasonable, but the BFBT analysis reveals a tendency of CTF to overpredict void, especially in the slug flow regime. The execution of these tests is fully automated, analysis is documented in the CTF Validation and Verification manual, and the tests have become part of CASL continuous regression testing system. This paper will summarize these recent developments and some of the two-phase assessments that have been performed on CTF.« less

OVERVIEW OF NEUTRON MEASUREMENTS IN JET FUSION DEVICE.

PubMed

Batistoni, P; Villari, R; Obryk, B; Packer, L W; Stamatelatos, I E; Popovichev, S; Colangeli, A; Colling, B; Fonnesu, N; Loreti, S; Klix, A; Klosowski, M; Malik, K; Naish, J; Pillon, M; Vasilopoulou, T; De Felice, P; Pimpinella, M; Quintieri, L

2017-10-05

The design and operation of ITER experimental fusion reactor requires the development of neutron measurement techniques and numerical tools to derive the fusion power and the radiation field in the device and in the surrounding areas. Nuclear analyses provide essential input to the conceptual design, optimisation, engineering and safety case in ITER and power plant studies. The required radiation transport calculations are extremely challenging because of the large physical extent of the reactor plant, the complexity of the geometry, and the combination of deep penetration and streaming paths. This article reports the experimental activities which are carried-out at JET to validate the neutronics measurements methods and numerical tools used in ITER and power plant design. A new deuterium-tritium campaign is proposed in 2019 at JET: the unique 14 MeV neutron yields produced will be exploited as much as possible to validate measurement techniques, codes, procedures and data currently used in ITER design thus reducing the related uncertainties and the associated risks in the machine operation. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

A 2-D/1-D transverse leakage approximation based on azimuthal, Fourier moments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stimpson, Shane G.; Collins, Benjamin S.; Downar, Thomas

Here, the MPACT code being developed collaboratively by Oak Ridge National Laboratory and the University of Michigan is the primary deterministic neutron transport solver within the Virtual Environment for Reactor Applications Core Simulator (VERA-CS). In MPACT, the two-dimensional (2-D)/one-dimensional (1-D) scheme is the most commonly used method for solving neutron transport-based three-dimensional nuclear reactor core physics problems. Several axial solvers in this scheme assume isotropic transverse leakages, but work with the axial S N solver has extended these leakages to include both polar and azimuthal dependence. However, explicit angular representation can be burdensome for run-time and memory requirements. The workmore » here alleviates this burden by assuming that the azimuthal dependence of the angular flux and transverse leakages are represented by a Fourier series expansion. At the heart of this is a new axial SN solver that takes in a Fourier expanded radial transverse leakage and generates the angular fluxes used to construct the axial transverse leakages used in the 2-D-Method of Characteristics calculations.« less

A 2-D/1-D transverse leakage approximation based on azimuthal, Fourier moments

DOE PAGES

Stimpson, Shane G.; Collins, Benjamin S.; Downar, Thomas

2017-01-12

Here, the MPACT code being developed collaboratively by Oak Ridge National Laboratory and the University of Michigan is the primary deterministic neutron transport solver within the Virtual Environment for Reactor Applications Core Simulator (VERA-CS). In MPACT, the two-dimensional (2-D)/one-dimensional (1-D) scheme is the most commonly used method for solving neutron transport-based three-dimensional nuclear reactor core physics problems. Several axial solvers in this scheme assume isotropic transverse leakages, but work with the axial S N solver has extended these leakages to include both polar and azimuthal dependence. However, explicit angular representation can be burdensome for run-time and memory requirements. The workmore » here alleviates this burden by assuming that the azimuthal dependence of the angular flux and transverse leakages are represented by a Fourier series expansion. At the heart of this is a new axial SN solver that takes in a Fourier expanded radial transverse leakage and generates the angular fluxes used to construct the axial transverse leakages used in the 2-D-Method of Characteristics calculations.« less

Integrated Tokamak modeling: When physics informs engineering and research planning

NASA Astrophysics Data System (ADS)

Poli, Francesca Maria

2018-05-01

Modeling tokamaks enables a deeper understanding of how to run and control our experiments and how to design stable and reliable reactors. We model tokamaks to understand the nonlinear dynamics of plasmas embedded in magnetic fields and contained by finite size, conducting structures, and the interplay between turbulence, magneto-hydrodynamic instabilities, and wave propagation. This tutorial guides through the components of a tokamak simulator, highlighting how high-fidelity simulations can guide the development of reduced models that can be used to understand how the dynamics at a small scale and short time scales affects macroscopic transport and global stability of plasmas. It discusses the important role that reduced models have in the modeling of an entire plasma discharge from startup to termination, the limits of these models, and how they can be improved. It discusses the important role that efficient workflows have in the coupling between codes, in the validation of models against experiments and in the verification of theoretical models. Finally, it reviews the status of integrated modeling and addresses the gaps and needs towards predictions of future devices and fusion reactors.

Integrated Tokamak modeling: When physics informs engineering and research planning

DOE Office of Scientific and Technical Information (OSTI.GOV)

Poli, Francesca Maria

Modeling tokamaks enables a deeper understanding of how to run and control our experiments and how to design stable and reliable reactors. We model tokamaks to understand the nonlinear dynamics of plasmas embedded in magnetic fields and contained by finite size, conducting structures, and the interplay between turbulence, magneto-hydrodynamic instabilities, and wave propagation. This tutorial guides through the components of a tokamak simulator, highlighting how high-fidelity simulations can guide the development of reduced models that can be used to understand how the dynamics at a small scale and short time scales affects macroscopic transport and global stability of plasmas. Itmore » discusses the important role that reduced models have in the modeling of an entire plasma discharge from startup to termination, the limits of these models, and how they can be improved. It discusses the important role that efficient workflows have in the coupling between codes, in the validation of models against experiments and in the verification of theoretical models. Finally, it reviews the status of integrated modeling and addresses the gaps and needs towards predictions of future devices and fusion reactors.« less

Integrated Tokamak modeling: When physics informs engineering and research planning

DOE PAGES

Poli, Francesca Maria

2018-05-01

Modeling tokamaks enables a deeper understanding of how to run and control our experiments and how to design stable and reliable reactors. We model tokamaks to understand the nonlinear dynamics of plasmas embedded in magnetic fields and contained by finite size, conducting structures, and the interplay between turbulence, magneto-hydrodynamic instabilities, and wave propagation. This tutorial guides through the components of a tokamak simulator, highlighting how high-fidelity simulations can guide the development of reduced models that can be used to understand how the dynamics at a small scale and short time scales affects macroscopic transport and global stability of plasmas. Itmore » discusses the important role that reduced models have in the modeling of an entire plasma discharge from startup to termination, the limits of these models, and how they can be improved. It discusses the important role that efficient workflows have in the coupling between codes, in the validation of models against experiments and in the verification of theoretical models. Finally, it reviews the status of integrated modeling and addresses the gaps and needs towards predictions of future devices and fusion reactors.« less

Fusion Safety Program annual report, fiscal year 1994

NASA Astrophysics Data System (ADS)

Longhurst, Glen R.; Cadwallader, Lee C.; Dolan, Thomas J.; Herring, J. Stephen; McCarthy, Kathryn A.; Merrill, Brad J.; Motloch, Chester C.; Petti, David A.

1995-03-01

This report summarizes the major activities of the Fusion Safety Program in fiscal year 1994. The Idaho National Engineering Laboratory (INEL) is the designated lead laboratory and Lockheed Idaho Technologies Company is the prime contractor for this program. The Fusion Safety Program was initiated in 1979. Activities are conducted at the INEL, at other DOE laboratories, and at other institutions, including the University of Wisconsin. The technical areas covered in this report include tritium safety, beryllium safety, chemical reactions and activation product release, safety aspects of fusion magnet systems, plasma disruptions, risk assessment failure rate data base development, and thermalhydraulics code development and their application to fusion safety issues. Much of this work has been done in support of the International Thermonuclear Experimental Reactor (ITER). Also included in the report are summaries of the safety and environmental studies performed by the Fusion Safety Program for the Tokamak Physics Experiment and the Tokamak Fusion Test Reactor and of the technical support for commercial fusion facility conceptual design studies. A major activity this year has been work to develop a DOE Technical Standard for the safety of fusion test facilities.

A Numerical Study of the Non-Ideal Behavior, Parameters, and Novel Applications of an Electrothermal Plasma Source

NASA Astrophysics Data System (ADS)

Winfrey, A. Leigh

Electrothermal plasma sources have numerous applications including hypervelocity launchers, fusion reactor pellet injection, and space propulsion systems. The time evolution of important plasma parameters at the source exit is important in determining the suitability of the source for different applications. In this study a capillary discharge code has been modified to incorporate non-ideal behavior by using an exact analytical model for the Coulomb logarithm in the plasma electrical conductivity formula. Actual discharge currents from electrothermal plasma experiments were used and code results for both ideal and non-ideal plasma models were compared to experimental data, specifically the ablated mass from the capillary and the electrical conductivity as measured by the discharge current and the voltage. Electrothermal plasma sources operating in the ablation-controlled arc regime use discharge currents with pulse lengths between 100 micros to 1 ms. Faster or longer or extended flat-top pulses can also be generated to satisfy various applications of ET sources. Extension of the peak current for up to an additional 1000 micros was tested. Calculations for non-ideal and ideal plasma models show that extended flattop pulses produce more ablated mass, which scales linearly with increased pulse length while other parameters remain almost constant. A new configuration of the PIPE source has been proposed in order to investigate the formation of plasmas from mixed materials. The electrothermal segmented plasma source can be used for studies related to surface coatings, surface modification, ion implantation, materials synthesis, and the physics of complex mixed plasmas. This source is a capillary discharge where the ablation liner is made from segments of different materials instead of a single sleeve. This system should allow for the modeling and characterization of the growth plasma as it provides all materials needed for fabrication through the same method. An ablation-free capillary discharge computer code has been developed to model plasma flow and acceleration of pellets for fusion fueling in magnetic fusion reactors. Two case studies with and without ablation, including different source configurations have been studied here. Velocities necessary for fusion fueling have been achieved. New additions made to the code model incorporate radial heat and energy transfer and move ETFLOW towards being a 2-D model of the plasma flow. This semi 2-D approach gives a view of the behavior of the plasma inside the capillary as it is affected by important physical parameters such as radial thermal heat conduction and their effect on wall ablation.

Dynamic characteristics of a VK-50 reactor operating under conditions of the loss of a normal feedwater flow

NASA Astrophysics Data System (ADS)

Semidotskiy, I. I.; Kurskiy, A. S.

2013-12-01

The paper describes the conditions of the ATWS type with virtually complete cessation of the feed-water flow at the operating power level of a reactor of the VK-50 type. Under these conditions, the role of spatial kinetics in the system of feedback between thermohydraulic and nuclear processes with bulk boiling of the coolant in the reactor core is clearly seen. This feature determines the specific character of experimental data obtained and the suitability of their use for verification of the associated codes used for calculating water-water reactors.

78 FR 2694 - Meeting of the ACRS Subcommittee on Planning and Procedures; Notice of Meeting

Federal Register 2010, 2011, 2012, 2013, 2014

2013-01-14

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Planning and Procedures; Notice of Meeting The ACRS Subcommittee on Planning and Procedures..., Advisory Committee on Reactor Safeguards. [FR Doc. 2013-00543 Filed 1-11-13; 8:45 am] BILLING CODE 7590-01...

141. ARAIII Equipment location plan. Includes list of equipment and ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

141. ARA-III Equipment location plan. Includes list of equipment and location in reactor, control, and other buildings. Aerojet-general 880-area/GCRE-101-U-1. Date: February 1958. Ineel index code no. 063-0101-65-013-192508. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

75 FR 16517 - FirstEnergy Nuclear Operating Company; Environmental Assessment and Finding of No Significant Impact

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-01

... Perry. FOR FURTHER INFORMATION CONTACT: Michael Mahoney, Office of Nuclear Reactor Regulation, U.S... Nuclear Reactor Regulation. [FR Doc. 2010-7331 Filed 3-31-10; 8:45 am] BILLING CODE 7590-01-P ... NUCLEAR REGULATORY COMMISSION [Docket No. 50-440; NRC-2010-0124] FirstEnergy Nuclear Operating...

Study of fusion product effects in field-reversed mirrors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Driemeyer, D.E.

1980-01-01

The effect of fusion products (fps) on Field-Reversed Mirror (FRM) reactor concepts has been evaluated through the development of two new computer models. The first code (MCFRM) treats fps as test particles in a fixed background plasma, which is represented as a fluid. MCFRM includes a Monte Carlo treatment of Coulomb scattering and thus provides an accurate treatment of fp behavior even at lower energies where pitch-angle scattering becomes important. The second code (FRMOD) is a steady-state, globally averaged, two-fluid (ion and electron), point model of the FRM plasma that incorporates fp heating and ash buildup values which are consistentmore » with the MCFRM calculations. These codes have been used extensively in the development of an advanced-fuel FRM reactor design (SAFFIRE). A Catalyzed-D version of the plant is also discussed along with an investigation of the steady-state energy distribution of fps in the FRM. User guides for the two computer codes are also included.« less

Development and application of the dynamic system doctor to nuclear reactor probabilistic risk assessments.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kunsman, David Marvin; Aldemir, Tunc; Rutt, Benjamin

2008-05-01

This LDRD project has produced a tool that makes probabilistic risk assessments (PRAs) of nuclear reactors - analyses which are very resource intensive - more efficient. PRAs of nuclear reactors are being increasingly relied on by the United States Nuclear Regulatory Commission (U.S.N.R.C.) for licensing decisions for current and advanced reactors. Yet, PRAs are produced much as they were 20 years ago. The work here applied a modern systems analysis technique to the accident progression analysis portion of the PRA; the technique was a system-independent multi-task computer driver routine. Initially, the objective of the work was to fuse the accidentmore » progression event tree (APET) portion of a PRA to the dynamic system doctor (DSD) created by Ohio State University. Instead, during the initial efforts, it was found that the DSD could be linked directly to a detailed accident progression phenomenological simulation code - the type on which APET construction and analysis relies, albeit indirectly - and thereby directly create and analyze the APET. The expanded DSD computational architecture and infrastructure that was created during this effort is called ADAPT (Analysis of Dynamic Accident Progression Trees). ADAPT is a system software infrastructure that supports execution and analysis of multiple dynamic event-tree simulations on distributed environments. A simulator abstraction layer was developed, and a generic driver was implemented for executing simulators on a distributed environment. As a demonstration of the use of the methodological tool, ADAPT was applied to quantify the likelihood of competing accident progression pathways occurring for a particular accident scenario in a particular reactor type using MELCOR, an integrated severe accident analysis code developed at Sandia. (ADAPT was intentionally created with flexibility, however, and is not limited to interacting with only one code. With minor coding changes to input files, ADAPT can be linked to other such codes.) The results of this demonstration indicate that the approach can significantly reduce the resources required for Level 2 PRAs. From the phenomenological viewpoint, ADAPT can also treat the associated epistemic and aleatory uncertainties. This methodology can also be used for analyses of other complex systems. Any complex system can be analyzed using ADAPT if the workings of that system can be displayed as an event tree, there is a computer code that simulates how those events could progress, and that simulator code has switches to turn on and off system events, phenomena, etc. Using and applying ADAPT to particular problems is not human independent. While the human resources for the creation and analysis of the accident progression are significantly decreased, knowledgeable analysts are still necessary for a given project to apply ADAPT successfully. This research and development effort has met its original goals and then exceeded them.« less

Dynamic event tree analysis with the SAS4A/SASSYS-1 safety analysis code

DOE PAGES

Jankovsky, Zachary K.; Denman, Matthew R.; Aldemir, Tunc

2018-02-02

The consequences of a transient in an advanced sodium-cooled fast reactor are difficult to capture with the traditional approach to probabilistic risk assessment (PRA). Numerous safety-relevant systems are passive and may have operational states that cannot be represented by binary success or failure. In addition, the specific order and timing of events may be crucial which necessitates the use of dynamic PRA tools such as ADAPT. The modifications to the SAS4A/SASSYS-1 sodium-cooled fast reactor safety analysis code for linking it to ADAPT to perform a dynamic PRA are described. A test case is used to demonstrate the linking process andmore » to illustrate the type of insights that may be gained with this process. Finally, newly-developed dynamic importance measures are used to assess the significance of reactor parameters/constituents on calculated consequences of initiating events.« less

Dynamic event tree analysis with the SAS4A/SASSYS-1 safety analysis code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jankovsky, Zachary K.; Denman, Matthew R.; Aldemir, Tunc

The consequences of a transient in an advanced sodium-cooled fast reactor are difficult to capture with the traditional approach to probabilistic risk assessment (PRA). Numerous safety-relevant systems are passive and may have operational states that cannot be represented by binary success or failure. In addition, the specific order and timing of events may be crucial which necessitates the use of dynamic PRA tools such as ADAPT. The modifications to the SAS4A/SASSYS-1 sodium-cooled fast reactor safety analysis code for linking it to ADAPT to perform a dynamic PRA are described. A test case is used to demonstrate the linking process andmore » to illustrate the type of insights that may be gained with this process. Finally, newly-developed dynamic importance measures are used to assess the significance of reactor parameters/constituents on calculated consequences of initiating events.« less

LBE water interaction in sub-critical reactors: First experimental and modelling results

NASA Astrophysics Data System (ADS)

Ciampichetti, A.; Agostini, P.; Benamati, G.; Bandini, G.; Pellini, D.; Forgione, N.; Oriolo, F.; Ambrosini, W.

2008-06-01

This paper concerns the study of the phenomena involved in the interaction between LBE and pressurised water which could occur in some hypothetical accidents in accelerator driven system type reactors. The LIFUS 5 facility was designed and built at ENEA-Brasimone to reproduce this kind of interaction in a wide range of conditions. The first test of the experimental program was carried out injecting water at 70 bar and 235 °C in a reaction vessel containing LBE at 1 bar and 350 °C. A pressurisation up to 80 bar was observed in the test section during the considered transient. The SIMMER III code was used to simulate the performed test. The calculated data agree in a satisfactory way with the experimental results giving confidence in the possibility to use this code for safety analyses of heavy liquid metal cooled reactors.

Burn Control Mechanisms in Tokamaks

NASA Astrophysics Data System (ADS)

Hill, M. A.; Stacey, W. M.

2015-11-01

Burn control and passive safety in accident scenarios will be an important design consideration in future tokamak reactors, in particular fusion-fission hybrid reactors, e.g. the Subcritical Advanced Burner Reactor. We are developing a burning plasma dynamics code to explore various aspects of burn control, with the intent to identify feedback mechanisms that would prevent power excursions. This code solves the coupled set of global density and temperature equations, using scaling relations from experimental fits. Predictions of densities and temperatures have been benchmarked against DIII-D data. We are examining several potential feedback mechanisms to limit power excursions: i) ion-orbit loss, ii) thermal instability density limits, iii) MHD instability limits, iv) the degradation of alpha-particle confinement, v) modifications to the radial current profile, vi) ``divertor choking'' and vii) Type 1 ELMs. Work supported by the US DOE under DE-FG02-00ER54538, DE-FC02-04ER54698.

Initial Neutronics Analyses for HEU to LEU Fuel Conversion of the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kontogeorgakos, D.; Derstine, K.; Wright, A.

2013-06-01

The purpose of the TREAT reactor is to generate large transient neutron pulses in test samples without over-heating the core to simulate fuel assembly accident conditions. The power transients in the present HEU core are inherently self-limiting such that the core prevents itself from overheating even in the event of a reactivity insertion accident. The objective of this study was to support the assessment of the feasibility of the TREAT core conversion based on the present reactor performance metrics and the technical specifications of the HEU core. The LEU fuel assembly studied had the same overall design, materials (UO 2more » particles finely dispersed in graphite) and impurities content as the HEU fuel assembly. The Monte Carlo N–Particle code (MCNP) and the point kinetics code TREKIN were used in the analyses.« less

Analysis of the SL-1 Accident Using RELAPS5-3D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Francisco, A.D. and Tomlinson, E. T.

2007-11-08

On January 3, 1961, at the National Reactor Testing Station, in Idaho Falls, Idaho, the Stationary Low Power Reactor No. 1 (SL-1) experienced a major nuclear excursion, killing three people, and destroying the reactor core. The SL-1 reactor, a 3 MW{sub t} boiling water reactor, was shut down and undergoing routine maintenance work at the time. This paper presents an analysis of the SL-1 reactor excursion using the RELAP5-3D thermal-hydraulic and nuclear analysis code, with the intent of simulating the accident from the point of reactivity insertion to destruction and vaporization of the fuel. Results are presented, along with amore » discussion of sensitivity to some reactor and transient parameters (many of the details are only known with a high level of uncertainty).« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boyack, B.E.

The PIUS reactor utilizes simplified, inherent, passive, or other innovative means to accomplish safety functions. Accordingly, the PIUS reactor is subject to the requirements of 10CFR52.47(b)(2)(i)(A). This regulation requires that the applicant adequately demonstrate the performance of each safety feature, interdependent effects among the safety features, and a sufficient data base on the safety features of the design to assess the analytical tools used for safety analysis. Los Alamos has assessed the quality and completeness of the existing and planned data bases used by Asea Brown Boveri to validate its safety analysis codes and other relevant data bases. Only amore » limited data base of separate effect and integral tests exist at present. This data base is not adequate to fulfill the requirements of 10CFR52.47(b)(2)(i)(A). Asea Brown Boveri has stated that it plans to conduct more separate effect and integral test programs. If appropriately designed and conducted, these test programs have the potential to satisfy most of the data base requirements of 10CFR52.47(b)(2)(i)(A) and remedy most of the deficiencies of the currently existing combined data base. However, the most important physical processes in PIUS are related to reactor shutdown because the PIUS reactor does not contain rodded shutdown and control systems. For safety-related reactor shutdown, PIUS relies on negative reactivity insertions from the moderator temperature coefficient and from boron entering the core from the reactor pool. Asea Brown Boveri has neither developed a direct experimental data base for these important processes nor provided a rationale for indirect testing of these key PIUS processes. This is assessed as a significant shortcoming. In preparing the conclusions of this report, test documentation and results have been reviewed for only one integral test program, the small-scale integral tests conducted in the ATLE facility.« less

ORIGEN-based Nuclear Fuel Inventory Module for Fuel Cycle Assessment: Final Project Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Skutnik, Steven E.

The goal of this project, “ORIGEN-based Nuclear Fuel Depletion Module for Fuel Cycle Assessment" is to create a physics-based reactor depletion and decay module for the Cyclus nuclear fuel cycle simulator in order to assess nuclear fuel inventories over a broad space of reactor operating conditions. The overall goal of this approach is to facilitate evaluations of nuclear fuel inventories for a broad space of scenarios, including extended used nuclear fuel storage and cascading impacts on fuel cycle options such as actinide recovery in used nuclear fuel, particularly for multiple recycle scenarios. The advantages of a physics-based approach (compared tomore » a recipe-based approach which has been typically employed for fuel cycle simulators) is in its inherent flexibility; such an approach can more readily accommodate the broad space of potential isotopic vectors that may be encountered under advanced fuel cycle options. In order to develop this flexible reactor analysis capability, we are leveraging the Origen nuclear fuel depletion and decay module from SCALE to produce a standalone “depletion engine” which will serve as the kernel of a Cyclus-based reactor analysis module. The ORIGEN depletion module is a rigorously benchmarked and extensively validated tool for nuclear fuel analysis and thus its incorporation into the Cyclus framework can bring these capabilities to bear on the problem of evaluating long-term impacts of fuel cycle option choices on relevant metrics of interest, including materials inventories and availability (for multiple recycle scenarios), long-term waste management and repository impacts, etc. Developing this Origen-based analysis capability for Cyclus requires the refinement of the Origen analysis sequence to the point where it can reasonably be compiled as a standalone sequence outside of SCALE; i.e., wherein all of the computational aspects of Origen (including reactor cross-section library processing and interpolation, input and output processing, and depletion/decay solvers) can be self-contained into a single executable sequence. Further, to embed this capability into other software environments (such as the Cyclus fuel cycle simulator) requires that Origen’s capabilities be encapsulated into a portable, self-contained library which other codes can then call directly through function calls, thereby directly accessing the solver and data processing capabilities of Origen. Additional components relevant to this work include modernization of the reactor data libraries used by Origen for conducting nuclear fuel depletion calculations. This work has included the development of new fuel assembly lattices not previously available (such as for CANDU heavy-water reactor assemblies) as well as validation of updated lattices for light-water reactors updated to employ modern nuclear data evaluations. The CyBORG reactor analysis module as-developed under this workscope is fully capable of dynamic calculation of depleted fuel compositions from all commercial U.S. reactor assembly types as well as a number of international fuel types, including MOX, VVER, MAGNOX, and PHWR CANDU fuel assemblies. In addition, the Origen-based depletion engine allows for CyBORG to evaluate novel fuel assembly and reactor design types via creation of Origen reactor data libraries via SCALE. The establishment of this new modeling capability affords fuel cycle modelers a substantially improved ability to model dynamically-changing fuel cycle and reactor conditions, including recycled fuel compositions from fuel cycle scenarios involving material recycle into thermal-spectrum systems.« less

Developments in Sensitivity Methodologies and the Validation of Reactor Physics Calculations

DOE PAGES

Palmiotti, Giuseppe; Salvatores, Massimo

2012-01-01

The sensitivity methodologies have been a remarkable story when adopted in the reactor physics field. Sensitivity coefficients can be used for different objectives like uncertainty estimates, design optimization, determination of target accuracy requirements, adjustment of input parameters, and evaluations of the representativity of an experiment with respect to a reference design configuration. A review of the methods used is provided, and several examples illustrate the success of the methodology in reactor physics. A new application as the improvement of nuclear basic parameters using integral experiments is also described.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Epiney, A.; Canepa, S.; Zerkak, O.

The STARS project at the Paul Scherrer Institut (PSI) has adopted the TRACE thermal-hydraulic (T-H) code for best-estimate system transient simulations of the Swiss Light Water Reactors (LWRs). For analyses involving interactions between system and core, a coupling of TRACE with the SIMULATE-3K (S3K) LWR core simulator has also been developed. In this configuration, the TRACE code and associated nuclear power reactor simulation models play a central role to achieve a comprehensive safety analysis capability. Thus, efforts have now been undertaken to consolidate the validation strategy by implementing a more rigorous and structured assessment approach for TRACE applications involving eithermore » only system T-H evaluations or requiring interfaces to e.g. detailed core or fuel behavior models. The first part of this paper presents the preliminary concepts of this validation strategy. The principle is to systematically track the evolution of a given set of predicted physical Quantities of Interest (QoIs) over a multidimensional parametric space where each of the dimensions represent the evolution of specific analysis aspects, including e.g. code version, transient specific simulation methodology and model "nodalisation". If properly set up, such environment should provide code developers and code users with persistent (less affected by user effect) and quantified information (sensitivity of QoIs) on the applicability of a simulation scheme (codes, input models, methodology) for steady state and transient analysis of full LWR systems. Through this, for each given transient/accident, critical paths of the validation process can be identified that could then translate into defining reference schemes to be applied for downstream predictive simulations. In order to illustrate this approach, the second part of this paper presents a first application of this validation strategy to an inadvertent blowdown event that occurred in a Swiss BWR/6. The transient was initiated by the spurious actuation of the Automatic Depressurization System (ADS). The validation approach progresses through a number of dimensions here: First, the same BWR system simulation model is assessed for different versions of the TRACE code, up to the most recent one. The second dimension is the "nodalisation" dimension, where changes to the input model are assessed. The third dimension is the "methodology" dimension. In this case imposed power and an updated TRACE core model are investigated. For each step in each validation dimension, a common set of QoIs are investigated. For the steady-state results, these include fuel temperatures distributions. For the transient part of the present study, the evaluated QoIs include the system pressure evolution and water carry-over into the steam line.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

Progress is reported on fundamental research in: crystal physics, reactions at metal surfaces, spectroscopy of ionic media, structure of metals, theory of alloying, physical properties, sintering, deformation of crystalline solids, x ray diffraction, metallurgy of superconducting materials, and electron microscope studies. Long-randge applied research studies were conducted for: zirconium metallurgy, materials compatibility, solid reactions, fuel element development, mechanical properties, non-destructive testing, and high-temperature materials. Reactor development support work was carried out for: gas-cooled reactor program, molten-salt reactor, high-flux isotope reactor, space-power program, thorium-utilization program, advanced-test reactor, Army Package Power Reactor, Enrico Fermi fast-breeder reactor, and water desalination program. Other programmore » activities, for which research was conducted, included: thermonuclear project, transuraniunn program, and post-irradiation examination laboratory. Separate abstracts were prepared for 30 sections of the report. (B.O.G.)« less

Modeling of two-phase flow instabilities during startup transients utilizing RAMONA-4B methodology

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paniagua, J.; Rohatgi, U.S.; Prasad, V.

1996-10-01

RAMONA-4B code is currently under development for simulating thermal hydraulic instabilities that can occur in Boiling Water Reactors (BWRs) and the Simplified Boiling Water Reactor (SBWR). As one of the missions of RAMONA-4B is to simulate SBWR startup transients, where geysering or condensation-induced instability may be encountered, the code needs to be assessed for this application. This paper outlines the results of the assessments of the current version of RAMONA-4B and the modifications necessary for simulating the geysering or condensation-induced instability. The test selected for assessment are the geysering tests performed by Prof Aritomi (1993).

Mesos-scale modeling of irradiation in pressurized water reactor concrete biological shields

DOE Office of Scientific and Technical Information (OSTI.GOV)

Le Pape, Yann; Huang, Hai

Neutron irradiation exposure causes aggregate expansion, namely radiation-induced volumetric expansion (RIVE). The structural significance of RIVE on a portion of a prototypical pressurized water reactor (PWR) concrete biological shield (CBS) is investigated by using a meso- scale nonlinear concrete model with inputs from an irradiation transport code and a coupled moisture transport-heat transfer code. RIVE-induced severe cracking onset appears to be triggered by the ini- tial shrinkage-induced cracking and propagates to a depth of > 10 cm at extended operation of 80 years. Relaxation of the cement paste stresses results in delaying the crack propagation by about 10 years.

TOWARD THE DEVELOPMENT OF A CONSENSUS MATERIALS DATABASE FOR PRESSURE TECHNOLGY APPLICATIONS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swindeman, Robert W; Ren, Weiju

The ASME construction code books specify materials and fabrication procedures that are acceptable for pressure technology applications. However, with few exceptions, the materials properties provided in the ASME code books provide no statistics or other information pertaining to material variability. Such information is central to the prediction and prevention of failure events. Many sources of materials data exist that provide variability information but such sources do not necessarily represent a consensus of experts with respect to the reported trends that are represented. Such a need has been identified by the ASME Standards Technology, LLC and initial steps have been takenmore » to address these needs: however, these steps are limited to project-specific applications only, such as the joint DOE-ASME project on materials for Generation IV nuclear reactors. In contrast to light-water reactor technology, the experience base for the Generation IV nuclear reactors is somewhat lacking and heavy reliance must be placed on model development and predictive capability. The database for model development is being assembled and includes existing code alloys such as alloy 800H and 9Cr-1Mo-V steel. Ownership and use rights are potential barriers that must be addressed.« less

Structural Integrity of Water Reactor Pressure Boundary Components.

DTIC Science & Technology

1980-08-01

Boiler and Pressure Vessel Code , Sec. Ill). Estimates of the upper shelf K level from small-specimen...from Appendix A of Section XI of the ASME Boiler and Pressure Vessel Code [11. Figure 9 shows this same data set, together with earlier data for...0969, NRL Memo- randum Report 4063, Sep. 1979. 11. Section XI - ASME Boiler and Pressure Vessel Code , Rules for Inservice Inspection of Nuclear

10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... nonpower reactors licensed to operate at or above a power level of 2 megawatts thermal. [38 FR 35430, Dec... OF PLANTS AND MATERIALS Physical Protection Requirements at Fixed Sites § 73.60 Additional...

10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... nonpower reactors licensed to operate at or above a power level of 2 megawatts thermal. [38 FR 35430, Dec... OF PLANTS AND MATERIALS Physical Protection Requirements at Fixed Sites § 73.60 Additional...

10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... nonpower reactors licensed to operate at or above a power level of 2 megawatts thermal. [38 FR 35430, Dec... OF PLANTS AND MATERIALS Physical Protection Requirements at Fixed Sites § 73.60 Additional...

Neutron fluxes in test reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Youinou, Gilles Jean-Michel

Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

Fatigue Behavior of HY-130 Steel Weldments Containing Fabrication Discontinuities.

DTIC Science & Technology

1985-04-18

discontinuities to solutions for elliptical discontinuities. One such approach has been formalized in the ASME Section XI Boiler and Pressure Vessel Code [1... Boiler and Pressure Vessel Code , Section XI, "Rules for Inservice Inspection of Nuclear Reactor Coolant Systems," American Society of Mechanical

Thermoelectric pump performance analysis computer code

NASA Technical Reports Server (NTRS)

Johnson, J. L.

1973-01-01

A computer program is presented that was used to analyze and design dual-throat electromagnetic dc conduction pumps for the 5-kwe ZrH reactor thermoelectric system. In addition to a listing of the code and corresponding identification of symbols, the bases for this analytical model are provided.

Coupling the System Analysis Module with SAS4A/SASSYS-1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fanning, T. H.; Hu, R.

2016-09-30

SAS4A/SASSYS-1 is a simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced reactors, with an emphasis on sodium fast reactors. SAS4A/SASSYS-1 has been under development and in active use for nearly forty-five years, and is currently maintained by the U.S. Department of Energy under the Office of Advanced Reactor Technology. Although SAS4A/SASSYS-1 contains a very capable primary and intermediate system modeling component, PRIMAR-4, it also has some shortcomings: outdated data management and code structure makes extension of the PRIMAR-4 module somewhat difficult. The user input format formore » PRIMAR-4 also limits the number of volumes and segments that can be used to describe a given system. The System Analysis Module (SAM) is a fairly new code development effort being carried out under the U.S. DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. SAM is being developed with advanced physical models, numerical methods, and software engineering practices; however, it is currently somewhat limited in the system components and phenomena that can be represented. For example, component models for electromagnetic pumps and multi-layer stratified volumes have not yet been developed. Nor is there support for a balance of plant model. Similarly, system-level phenomena such as control-rod driveline expansion and vessel elongation are not represented. This report documents fiscal year 2016 work that was carried out to couple the transient safety analysis capabilities of SAS4A/SASSYS-1 with the system modeling capabilities of SAM under the joint support of the ART and NEAMS programs. The coupling effort was successful and is demonstrated by evaluating an unprotected loss of flow transient for the Advanced Burner Test Reactor (ABTR) design. There are differences between the stand-alone SAS4A/SASSYS-1 simulations and the coupled SAS/SAM simulations, but these are mainly attributed to the limited maturity of the SAM development effort. The severe accident modeling capabilities in SAS4A/SASSYS-1 (sodium boiling, fuel melting and relocation) will continue to play a vital role for a long time. Therefore, the SAS4A/SASSYS-1 modernization effort should remain a high priority task under the ART program to ensure continued participation in domestic and international SFR safety collaborations and design optimizations. On the other hand, SAM provides an advanced system analysis tool, with improved numerical solution schemes, data management, code flexibility, and accuracy. SAM is still in early stages of development and will require continued support from NEAMS to fulfill its potential and to mature into a production tool for advanced reactor safety analysis. The effort to couple SAS4A/SASSYS-1 and SAM is the first step on the integration of these modeling capabilities.« less

MCNP-REN - A Monte Carlo Tool for Neutron Detector Design Without Using the Point Model

DOE Office of Scientific and Technical Information (OSTI.GOV)

Abhold, M.E.; Baker, M.C.

1999-07-25

The development of neutron detectors makes extensive use of the predictions of detector response through the use of Monte Carlo techniques in conjunction with the point reactor model. Unfortunately, the point reactor model fails to accurately predict detector response in common applications. For this reason, the general Monte Carlo N-Particle code (MCNP) was modified to simulate the pulse streams that would be generated by a neutron detector and normally analyzed by a shift register. This modified code, MCNP - Random Exponentially Distributed Neutron Source (MCNP-REN), along with the Time Analysis Program (TAP) predict neutron detector response without using the pointmore » reactor model, making it unnecessary for the user to decide whether or not the assumptions of the point model are met for their application. MCNP-REN is capable of simulating standard neutron coincidence counting as well as neutron multiplicity counting. Measurements of MOX fresh fuel made using the Underwater Coincidence Counter (UWCC) as well as measurements of HEU reactor fuel using the active neutron Research Reactor Fuel Counter (RRFC) are compared with calculations. The method used in MCNP-REN is demonstrated to be fundamentally sound and shown to eliminate the need to use the point model for detector performance predictions.« less

A Programmable Liquid Collimator for Both Coded Aperture Adaptive Imaging and Multiplexed Compton Scatter Tomography

DTIC Science & Technology

2012-03-01

environments where a source is either weak or shielded. A vehicle of this type could survey large areas after a nuclear attack or a nuclear reactor accident...to prevent its detection by γ-rays. The best application for unmanned vehicles is the detection of radioactive material after a nuclear reactor ...accident or a nuclear weapon detonation [70]. Whether by a nuclear detonation or a nuclear reactor accident, highly radioactive substances could be dis

Methodology, status, and plans for development and assessment of the RELAP5 code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johnson, G.W.; Riemke, R.A.

1997-07-01

RELAP/MOD3 is a computer code used for the simulation of transients and accidents in light-water nuclear power plants. The objective of the program to develop and maintain RELAP5 was and is to provide the U.S. Nuclear Regulatory Commission with an independent tool for assessing reactor safety. This paper describes code requirements, models, solution scheme, language and structure, user interface validation, and documentation. The paper also describes the current and near term development program and provides an assessment of the code`s strengths and limitations.

The behaviour of transuranic mixed oxide fuel in a Candu-900 reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Morreale, A. C.; Ball, M. R.; Novog, D. R.

2012-07-01

The production of transuranic actinide fuels for use in current thermal reactors provides a useful intermediary step in closing the nuclear fuel cycle. Extraction of actinides reduces the longevity, radiation and heat loads of spent material. The burning of transuranic fuels in current reactors for a limited amount of cycles reduces the infrastructure demand for fast reactors and provides an effective synergy that can result in a reduction of as much as 95% of spent fuel waste while reducing the fast reactor infrastructure needed by a factor of almost 13.5 [1]. This paper examines the features of actinide mixed oxidemore » fuel, TRUMOX, in a CANDU{sup R}* nuclear reactor. The actinide concentrations used were based on extraction from 30 year cooled spent fuel and mixed with natural uranium in 3.1 wt% actinide MOX fuel. Full lattice cell modeling was performed using the WIMS-AECL code, super-cell calculations were analyzed in DRAGON and full core analysis was executed in the RFSP 2-group diffusion code. A time-average full core model was produced and analyzed for reactor coefficients, reactivity device worth and online fuelling impacts. The standard CANDU operational limits were maintained throughout operations. The TRUMOX fuel design achieved a burnup of 27.36 MWd/kg HE. A full TRUMOX fuelled CANDU was shown to operate within acceptable limits and provided a viable intermediary step for burning actinides. The recycling, reprocessing and reuse of spent fuels produces a much more sustainable and efficient nuclear fuel cycle. (authors)« less

76 FR 26320 - Entergy Operations, Inc.; Biweekly Notice; Notice of Issuance of Amendment to Facility Operating...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-05-06

... Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone (301..., Office of Nuclear Reactor Regulation. [FR Doc. 2011-11107 Filed 5-5-11; 8:45 am] BILLING CODE 7590-01-P ... NUCLEAR REGULATORY COMMISSION [NRC-2011-0071; Docket No. 50-382] Entergy Operations, Inc...

76 FR 13676 - Exelon Generation Company, LLC; Notice of Withdrawal of Application for Amendment to Facility...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-03-14

.... Brown, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555..., Office of Nuclear Reactor Regulation. [FR Doc. 2011-5756 Filed 3-11-11; 8:45 am] BILLING CODE 7590-01-P ... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-373 and 50-374; NRC-2011-0051] Exelon Generation...

75 FR 57535 - Northern States Power Company-Minnesota Notice of Issuance of Amendments to Facility Operating...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-21

... issuance. FOR FURTHER INFORMATION CONTACT: Thomas J. Wengert, Office of Nuclear Reactor Regulation, U.S... Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2010-23516 Filed 9-20-10; 8:45 am] BILLING CODE... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-282 and 50-306; NRC-2010-0290] Northern States Power...

78 FR 17450 - Entergy Nuclear Operations, Inc.; Entergy Operations, Inc.; Biweekly Notice; Notice of Issuance...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-21

... INFORMATION CONTACT: Nageswaran Kalyanam, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory..., Office of Nuclear Reactor Regulation. [FR Doc. 2013-06510 Filed 3-20-13; 8:45 am] BILLING CODE 7590-01-P ... NUCLEAR REGULATORY COMMISSION [NRC-2013-0012] [Docket Nos. 50-458, 50-155, 72-043, 50-003, 50-247...

Nuclear Engineering Computer Modules: Reactor Dynamics, RD-1 and RD-2.

ERIC Educational Resources Information Center

Onega, Ronald J.

The objective of the Reactor Dynamics Module, RD-1, is to obtain the kinetics equation without feedback and solve the kinetics equations numerically for one to six delayed neutron groups for time varying reactivity insertions. The computer code FUMOKI (Fundamental Mode Kinetics) will calculate the power as a function of time for either uranium or…

Development of the MCNPX depletion capability: A Monte Carlo linked depletion method that automates the coupling between MCNPX and CINDER90 for high fidelity burnup calculations

NASA Astrophysics Data System (ADS)

Fensin, Michael Lorne

Monte Carlo-linked depletion methods have gained recent interest due to the ability to more accurately model complex 3-dimesional geometries and better track the evolution of temporal nuclide inventory by simulating the actual physical process utilizing continuous energy coefficients. The integration of CINDER90 into the MCNPX Monte Carlo radiation transport code provides a high-fidelity completely self-contained Monte-Carlo-linked depletion capability in a well established, widely accepted Monte Carlo radiation transport code that is compatible with most nuclear criticality (KCODE) particle tracking features in MCNPX. MCNPX depletion tracks all necessary reaction rates and follows as many isotopes as cross section data permits in order to achieve a highly accurate temporal nuclide inventory solution. This work chronicles relevant nuclear history, surveys current methodologies of depletion theory, details the methodology in applied MCNPX and provides benchmark results for three independent OECD/NEA benchmarks. Relevant nuclear history, from the Oklo reactor two billion years ago to the current major United States nuclear fuel cycle development programs, is addressed in order to supply the motivation for the development of this technology. A survey of current reaction rate and temporal nuclide inventory techniques is then provided to offer justification for the depletion strategy applied within MCNPX. The MCNPX depletion strategy is then dissected and each code feature is detailed chronicling the methodology development from the original linking of MONTEBURNS and MCNP to the most recent public release of the integrated capability (MCNPX 2.6.F). Calculation results of the OECD/NEA Phase IB benchmark, H. B. Robinson benchmark and OECD/NEA Phase IVB are then provided. The acceptable results of these calculations offer sufficient confidence in the predictive capability of the MCNPX depletion method. This capability sets up a significant foundation, in a well established and supported radiation transport code, for further development of a Monte Carlo-linked depletion methodology which is essential to the future development of advanced reactor technologies that exceed the limitations of current deterministic based methods.

Characterization of gamma field in the JSI TRIGA reactor

NASA Astrophysics Data System (ADS)

Ambrožič, Klemen; Radulović, Vladimir; Snoj, Luka; Gruel, Adrien; Guillou, Mael Le; Blaise, Patrick; Destouches, Christophe; Barbot, Loïc

2018-01-01

Research reactors such as the "Jožzef Stefan" Institute TRIGA reactor have primarily been designed for experimentation and sample irradiation with neutrons. However recent developments in incorporating additional instrumentation for nuclear power plant support and with novel high flux material testing reactor designs, γ field characterization has become of great interest for the characterization of the changes in operational parameters of electronic devices and for the evaluation of γ heating of MTR's structural materials in a representative reactor Γ spectrum. In this paper, we present ongoing work on γ field characterization both experimentally, by performing γ field measurements, and by simulations, using Monte Carlo particle transport codes in conjunction with R2S methodology for delayed γ field characterization.

Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

NASA Technical Reports Server (NTRS)

Roth, R. J.

1976-01-01

The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

1978-02-01

The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocitymore » and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.« less

FY16 Status Report on NEAMS Neutronics Activities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, C. H.; Shemon, E. R.; Smith, M. A.

2016-09-30

The goal of the NEAMS neutronics effort is to develop a neutronics toolkit for use on sodium-cooled fast reactors (SFRs) which can be extended to other reactor types. The neutronics toolkit includes the high-fidelity deterministic neutron transport code PROTEUS and many supporting tools such as a cross section generation code MC 2-3, a cross section library generation code, alternative cross section generation tools, mesh generation and conversion utilities, and an automated regression test tool. The FY16 effort for NEAMS neutronics focused on supporting the release of the SHARP toolkit and existing and new users, continuing to develop PROTEUS functions necessarymore » for performance improvement as well as the SHARP release, verifying PROTEUS against available existing benchmark problems, and developing new benchmark problems as needed. The FY16 research effort was focused on further updates of PROTEUS-SN and PROTEUS-MOCEX and cross section generation capabilities as needed.« less

A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

NASA Astrophysics Data System (ADS)

Reed, Mark; Parker, Ronald R.; Forget, Benoit

2012-06-01

This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more proliferation-resistant than that bred by conventional fast reactors. Furthermore, it can maintain constant total hybrid power output as burnup proceeds by varying the neutron source strength.

Hardening neutron spectrum for advanced actinide transmutation experiments in the ATR.

PubMed

Chang, G S; Ambrosek, R G

2005-01-01

The most effective method for transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products is in a fast neutron spectrum reactor. In the absence of a fast test reactor in the United States, initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. Such a test facility, with a spectrum similar but somewhat softer than that of the liquid-metal fast breeder reactor (LMFBR), has been constructed in the INEEL's Advanced Test Reactor (ATR). The radial fission power distribution of the actinide fuel pin, which is an important parameter in fission gas release modelling, needs to be accurately predicted and the hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum is compared. The comparison analyses in this study are performed using MCWO, a well-developed tool that couples the Monte Carlo transport code MCNP with the isotope depletion and build-up code ORIGEN-2. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations and detailed radial fission power profile calculations for a typical fast reactor (LMFBR) neutron spectrum and the hardened neutron spectrum test region in the ATR. The MCWO-calculated results indicate that the cadmium basket used in the advanced fuel test assembly in the ATR can effectively depress the linear heat generation rate in the experimental fuels and harden the neutron spectrum in the test region.

Analysis of typical WWER-1000 severe accident scenarios

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sorokin, Yu.S.; Shchekoldin, V.V.; Borisov, L.N.

2004-07-01

At present in EDO 'Gidropress' there is a certain experience of performing the analyses of severe accidents of reactor plant with WWER with application of domestic and foreign codes. Important data were also obtained by the results of calculation modeling of integrated experiments with fuel assembly melting comprising a real fuel. Systematization and consideration of these data in development and assimilation of codes are extremely important in connection with large uncertainty still existing in understanding and adequate description of phenomenology of severe accidents. The presented report gives a comparison of analysis results of severe accidents of reactor plant with WWER-1000more » for two typical scenarios made by using American MELCOR code and the Russian RATEG/SVECHA/HEFEST code. The results of calculation modeling are compared using above codes with the data of experiment FPT1 with fuel assembly melting comprising a real fuel, which has been carried out at the facility Phebus (France). The obtained results are considered in the report from the viewpoint of: - adequacy of results of calculation modeling of separate phenomena during severe accidents of RP with WWER by using the above codes; - influence of uncertainties (degree of details of calculation models, choice of parameters of models etc.); - choice of those or other setup variables (options) in the used codes; - necessity of detailed modeling of processes and phenomena as applied to design justification of safety of RP with WWER. (authors)« less

CTF Theory Manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Avramova, Maria N.; Salko, Robert K.

Coolant-Boiling in Rod Arrays|Two Fluids (COBRA-TF) is a thermal/ hydraulic (T/H) simulation code designed for light water reactor (LWR) vessel analysis. It uses a two-fluid, three-field (i.e. fluid film, fluid drops, and vapor) modeling approach. Both sub-channel and 3D Cartesian forms of 9 conservation equations are available for LWR modeling. The code was originally developed by Pacific Northwest Laboratory in 1980 and had been used and modified by several institutions over the last few decades. COBRA-TF also found use at the Pennsylvania State University (PSU) by the Reactor Dynamics and Fuel Management Group (RDFMG) and has been improved, updated, andmore » subsequently re-branded as CTF. As part of the improvement process, it was necessary to generate sufficient documentation for the open-source code which had lacked such material upon being adopted by RDFMG. This document serves mainly as a theory manual for CTF, detailing the many two-phase heat transfer, drag, and important accident scenario models contained in the code as well as the numerical solution process utilized. Coding of the models is also discussed, all with consideration for updates that have been made when transitioning from COBRA-TF to CTF. Further documentation outside of this manual is also available at RDFMG which focus on code input deck generation and source code global variable and module listings.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Downar, Thomas

This report summarizes the current status of VERA-CS Verification and Validation for PWR Core Follow operation and proposes a multi-phase plan for continuing VERA-CS V&V in FY17 and FY18. The proposed plan recognizes the hierarchical nature of a multi-physics code system such as VERA-CS and the importance of first achieving an acceptable level of V&V on each of the single physics codes before focusing on the V&V of the coupled physics solution. The report summarizes the V&V of each of the single physics codes systems currently used for core follow analysis (ie MPACT, CTF, Multigroup Cross Section Generation, and BISONmore » / Fuel Temperature Tables) and proposes specific actions to achieve a uniformly acceptable level of V&V in FY17. The report also recognizes the ongoing development of other codes important for PWR Core Follow (e.g. TIAMAT, MAMBA3D) and proposes Phase II (FY18) VERA-CS V&V activities in which those codes will also reach an acceptable level of V&V. The report then summarizes the current status of VERA-CS multi-physics V&V for PWR Core Follow and the ongoing PWR Core Follow V&V activities for FY17. An automated procedure and output data format is proposed for standardizing the output for core follow calculations and automatically generating tables and figures for the VERA-CS Latex file. A set of acceptance metrics is also proposed for the evaluation and assessment of core follow results that would be used within the script to automatically flag any results which require further analysis or more detailed explanation prior to being added to the VERA-CS validation base. After the Automation Scripts have been completed and tested using BEAVRS, the VERA-CS plan proposes the Watts Bar cycle depletion cases should be performed with the new cross section library and be included in the first draft of the new VERA-CS manual for release at the end of PoR15. Also, within the constraints imposed by the proprietary nature of plant data, as many as possible of the FY17 AMA Plant Core Follow cases should also be included in the VERA-CS manual at the end of PoR15. After completion of the ongoing development of TIAMAT for fully coupled, full core calculations with VERA-CS / BISON 1.5D, and after the completion of the refactoring of MAMBA3D for CIPS analysis in FY17, selected cases from the VERA-CS validation based should be performed, beginning with the legacy cases of Watts Bar and BEAVRS in PoR16. Finally, as potential Phase III future work some additional considerations are identified for extending the VERA-CS V&V to other reactor types such as the BWR.« less

Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jensen, C.; Wachs, D.; Carmack, J.

The restart of the Transient REActor Test (TREAT) facility provides a unique opportunity to engage the fast reactor fuels community to reinitiate in-pile experimental safety studies. Historically, the TREAT facility played a critical role in characterizing the behavior of both metal and oxide fast reactor fuels under off-normal conditions, irradiating hundreds of fuel pins to support fast reactor fuel development programs. The resulting test data has provided validation for a multitude of fuel performance and severe accident analysis computer codes. This paper will provide a review of the historical database of TREAT experiments including experiment design, instrumentation, test objectives, andmore » salient findings. Additionally, the paper will provide an introduction to the current and future experiment plans of the U.S. transient testing program at TREAT.« less

78 FR 69139 - Physical Security-Design Certification and Operating Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-18

... scheduled to close on October 30, 2013. The Nuclear Energy Institute (NEI) submitted a letter on October 9... NUCLEAR REGULATORY COMMISSION [NRC-2013-0225] Physical Security--Design Certification and Operating Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Standard review plan--draft section...

Conceptual study of an ICRH traveling-wave antenna system for low-coupling conditions as expected in DEMO

NASA Astrophysics Data System (ADS)

Ragona, R.; Messiaen, A.

2016-07-01

For the central heating of a fusion reactor ion cyclotron radio frequency heating (ICRH) is the first choice method as it is able to couple RF power to the ions without density limit. The drawback of this heating method is the problem of excitation of the magneto-sonic wave through the plasma boundary layer from the antenna located along the wall, without exceeding its voltage standoff. The amount of coupling depends on the antenna excitation and the surface admittance at the antenna output due to the plasma profile. The paper deals with the optimization of the antenna excitation by the use of sections of traveling-wave antennas (TWAs) distributed all along the reactor wall between the blanket modules. They are mounted and fed in resonant ring system(s). First, the physics of the coupling of a strap array is studied by simple models and the coupling code ANTITER II. Then, after the study of the basic properties of a TWA section, its feeding problem is solved by hybrids driving them in resonant ring circuit(s). The complete modeling is obtained from the matrices of the TWA sections connected to one of the feeding hybrid(s). The solution is iterated with the coupling code to determine the loading for a reference low-coupling ITER plasma profile. The resulting wave pattern up to the plasma bulk is derived. The proposed system is totally load resilient and allows us to obtain a very selective exciting wave spectrum. A discussion of some practical implementation problems is added.

The development of fuel performance models at the European institute for transuranium elements

NASA Astrophysics Data System (ADS)

Lassmann, K.; Ronchi, C.; Small, G. J.

1989-07-01

The design and operational performance of fuel rods for nuclear power stations has been the subject of detailed experimental research for over thirty years. In the last two decades the continuous demands for greater economy in conjunction with more stringent safety criteria have led to an increasing reliance on computer simulations. Conditions within a fuel rod must be calculated both for normal operation and for proposed reactor faults. It has thus been necessary to build up a reliable, theoretical understanding of the intricate physical, mechanical and chemical processes occurring under a wide range of conditions to obtain a quantitative insight into the behaviour of the fuel. A prime requirement, which has also proved to be the most taxing, is to predict the conditions under which failure of the cladding might occur, particularly in fuel nearing the end of its useful life. In this paper the general requirements of a fuel performance code are discussed briefly and an account is given of the basic concepts of code construction. An overview is then given of recent progress at the European Institute for Transuranium Elements in the development of a fuel rod performance code for general application and of more detailed mechanistic models for fission product behaviour.

Technical Basis for Physical Fidelity of NRC Control Room Training Simulators for Advanced Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Minsk, Brian S.; Branch, Kristi M.; Bates, Edward K.

2009-10-09

The objective of this study is to determine how simulator physical fidelity influences the effectiveness of training the regulatory personnel responsible for examination and oversight of operating personnel and inspection of technical systems at nuclear power reactors. It seeks to contribute to the U.S. Nuclear Regulatory Commission’s (NRC’s) understanding of the physical fidelity requirements of training simulators. The goal of the study is to provide an analytic framework, data, and analyses that inform NRC decisions about the physical fidelity requirements of the simulators it will need to train its staff for assignment at advanced reactors. These staff are expected tomore » come from increasingly diverse educational and experiential backgrounds.« less

Recent MELCOR and VICTORIA Fission Product Research at the NRC

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bixler, N.E.; Cole, R.K.; Gauntt, R.O.

1999-01-21

The MELCOR and VICTORIA severe accident analysis codes, which were developed at Sandia National Laboratories for the U. S. Nuclear Regulatory Commission, are designed to estimate fission product releases during nuclear reactor accidents in light water reactors. MELCOR is an integrated plant-assessment code that models the key phenomena in adequate detail for risk-assessment purposes. VICTORIA is a more specialized fission- product code that provides detailed modeling of chemical reactions and aerosol processes under the high-temperature conditions encountered in the reactor coolant system during a severe reactor accident. This paper focuses on recent enhancements and assessments of the two codes inmore » the area of fission product chemistry modeling. Recently, a model for iodine chemistry in aqueous pools in the containment building was incorporated into the MELCOR code. The model calculates dissolution of iodine into the pool and releases of organic and inorganic iodine vapors from the pool into the containment atmosphere. The main purpose of this model is to evaluate the effect of long-term revolatilization of dissolved iodine. Inputs to the model include dose rate in the pool, the amount of chloride-containing polymer, such as Hypalon, and the amount of buffering agents in the containment. Model predictions are compared against the Radioiodine Test Facility (RTF) experiments conduced by Atomic Energy of Canada Limited (AECL), specifically International Standard Problem 41. Improvements to VICTORIA's chemical reactions models were implemented as a result of recommendations from a peer review of VICTORIA that was completed last year. Specifically, an option is now included to model aerosols and deposited fission products as three condensed phases in addition to the original option of a single condensed phase. The three-condensed-phase model results in somewhat higher predicted fission product volatilities than does the single-condensed-phase model. Modeling of U02 thermochemistry was also improved, and results in better prediction of vaporization of uranium from fuel, which can react with released fission products to affect their volatility. This model also improves the prediction of fission product release rates from fuel. Finally, recent comparisons of MELCOR and VICTORIA with International Standard Problem 40 (STORM) data are presented. These comparisons focus on predicted therrnophoretic deposition, which is the dominant deposition mechanism. Sensitivity studies were performed with the codes to examine experimental and modeling uncertainties.« less

DE-NE0008277_PROTEUS final technical report 2018

DOE Office of Scientific and Technical Information (OSTI.GOV)

Enqvist, Andreas

This project details re-evaluations of experiments of gas-cooled fast reactor (GCFR) core designs performed in the 1970s at the PROTEUS reactor and create a series of International Reactor Physics Experiment Evaluation Project (IRPhEP) benchmarks. Currently there are no gas-cooled fast reactor (GCFR) experiments available in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). These experiments are excellent candidates for reanalysis and development of multiple benchmarks because these experiments provide high-quality integral nuclear data relevant to the validation and refinement of thorium, neptunium, uranium, plutonium, iron, and graphite cross sections. It would be cost prohibitive to reproduce suchmore » a comprehensive suite of experimental data to support any future GCFR endeavors.« less

126. ARAII Plot plan showing location of SL1 power plant ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

126. ARA-II Plot plan showing location of SL-1 power plant (reactor) building, and planned location of administrative and technical support building. C.A. Sundberg and Associates 866-area/ALPR-606-U-1. Date: May 1958. Ineel index code no. 070-0100-00-822-102834. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Establishment and assessment of code scaling capability

NASA Astrophysics Data System (ADS)

Lim, Jaehyok

In this thesis, a method for using RELAP5/MOD3.3 (Patch03) code models is described to establish and assess the code scaling capability and to corroborate the scaling methodology that has been used in the design of the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR applications (PUMA-E) facility. It was sponsored by the United States Nuclear Regulatory Commission (USNRC) under the program "PUMA ESBWR Tests". PUMA-E facility was built for the USNRC to obtain data on the performance of the passive safety systems of the General Electric (GE) Nuclear Energy Economic Simplified Boiling Water Reactor (ESBWR). Similarities between the prototype plant and the scaled-down test facility were investigated for a Gravity-Driven Cooling System (GDCS) Drain Line Break (GDLB). This thesis presents the results of the GDLB test, i.e., the GDLB test with one Isolation Condenser System (ICS) unit disabled. The test is a hypothetical multi-failure small break loss of coolant (SB LOCA) accident scenario in the ESBWR. The test results indicated that the blow-down phase, Automatic Depressurization System (ADS) actuation, and GDCS injection processes occurred as expected. The GDCS as an emergency core cooling system provided adequate supply of water to keep the Reactor Pressure Vessel (RPV) coolant level well above the Top of Active Fuel (TAF) during the entire GDLB transient. The long-term cooling phase, which is governed by the Passive Containment Cooling System (PCCS) condensation, kept the reactor containment system that is composed of Drywell (DW) and Wetwell (WW) below the design pressure of 414 kPa (60 psia). In addition, the ICS continued participating in heat removal during the long-term cooling phase. A general Code Scaling, Applicability, and Uncertainty (CSAU) evaluation approach was discussed in detail relative to safety analyses of Light Water Reactor (LWR). The major components of the CSAU methodology that were highlighted particularly focused on the scaling issues of experiments and models and their applicability to the nuclear power plant transient and accidents. The major thermal-hydraulic phenomena to be analyzed were identified and the predictive models adopted in RELAP5/MOD3.3 (Patch03) code were briefly reviewed.

Work plan for improving the DARWIN2.3 depleted material balance calculation of nuclides of interest for the fuel cycle

NASA Astrophysics Data System (ADS)

Rizzo, Axel; Vaglio-Gaudard, Claire; Martin, Julie-Fiona; Noguère, Gilles; Eschbach, Romain

2017-09-01

DARWIN2.3 is the reference package used for fuel cycle applications in France. It solves the Boltzmann and Bateman equations in a coupling way, with the European JEFF-3.1.1 nuclear data library, to compute the fuel cycle values of interest. It includes both deterministic transport codes APOLLO2 (for light water reactors) and ERANOS2 (for fast reactors), and the DARWIN/PEPIN2 depletion code, each of them being developed by CEA/DEN with the support of its industrial partners. The DARWIN2.3 package has been experimentally validated for pressurized and boiling water reactors, as well as for sodium fast reactors; this experimental validation relies on the analysis of post-irradiation experiments (PIE). The DARWIN2.3 experimental validation work points out some isotopes for which the depleted concentration calculation can be improved. Some other nuclides have no available experimental validation, and their concentration calculation uncertainty is provided by the propagation of a priori nuclear data uncertainties. This paper describes the work plan of studies initiated this year to improve the accuracy of the DARWIN2.3 depleted material balance calculation concerning some nuclides of interest for the fuel cycle.

Ex-vessel neutron dosimetry analysis for westinghouse 4-loop XL pressurized water reactor plant using the RadTrack{sup TM} Code System with the 3D parallel discrete ordinates code RAPTOR-M3G

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, J.; Alpan, F. A.; Fischer, G.A.

2011-07-01

Traditional two-dimensional (2D)/one-dimensional (1D) SYNTHESIS methodology has been widely used to calculate fast neutron (>1.0 MeV) fluence exposure to reactor pressure vessel in the belt-line region. However, it is expected that this methodology cannot provide accurate fast neutron fluence calculation at elevations far above or below the active core region. A three-dimensional (3D) parallel discrete ordinates calculation for ex-vessel neutron dosimetry on a Westinghouse 4-Loop XL Pressurized Water Reactor has been done. It shows good agreement between the calculated results and measured results. Furthermore, the results show very different fast neutron flux values at some of the former plate locationsmore » and elevations above and below an active core than those calculated by a 2D/1D SYNTHESIS method. This indicates that for certain irregular reactor internal structures, where the fast neutron flux has a very strong local effect, it is required to use a 3D transport method to calculate accurate fast neutron exposure. (authors)« less

Capabilities overview of the MORET 5 Monte Carlo code

NASA Astrophysics Data System (ADS)

Cochet, B.; Jinaphanh, A.; Heulers, L.; Jacquet, O.

2014-06-01

The MORET code is a simulation tool that solves the transport equation for neutrons using the Monte Carlo method. It allows users to model complex three-dimensional geometrical configurations, describe the materials, define their own tallies in order to analyse the results. The MORET code has been initially designed to perform calculations for criticality safety assessments. New features has been introduced in the MORET 5 code to expand its use for reactor applications. This paper presents an overview of the MORET 5 code capabilities, going through the description of materials, the geometry modelling, the transport simulation and the definition of the outputs.

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

Advances in boron neutron capture therapy (BNCT) at kyoto university - From reactor-based BNCT to accelerator-based BNCT

NASA Astrophysics Data System (ADS)

Sakurai, Yoshinori; Tanaka, Hiroki; Takata, Takushi; Fujimoto, Nozomi; Suzuki, Minoru; Masunaga, Shinichiro; Kinashi, Yuko; Kondo, Natsuko; Narabayashi, Masaru; Nakagawa, Yosuke; Watanabe, Tsubasa; Ono, Koji; Maruhashi, Akira

2015-07-01

At the Kyoto University Research Reactor Institute (KURRI), a clinical study of boron neutron capture therapy (BNCT) using a neutron irradiation facility installed at the research nuclear reactor has been regularly performed since February 1990. As of November 2014, 510 clinical irradiations were carried out using the reactor-based system. The world's first accelerator-based neutron irradiation system for BNCT clinical irradiation was completed at this institute in early 2009, and the clinical trial using this system was started in 2012. A shift of BCNT from special particle therapy to a general one is now in progress. To promote and support this shift, improvements to the irradiation system, as well as its preparation, and improvements in the physical engineering and the medical physics processes, such as dosimetry systems and quality assurance programs, must be considered. The recent advances in BNCT at KURRI are reported here with a focus on physical engineering and medical physics topics.

The Virtual Environment for Reactor Applications (VERA): Design and architecture

DOE Office of Scientific and Technical Information (OSTI.GOV)

Turner, John A., E-mail: turnerja@ornl.gov; Clarno, Kevin; Sieger, Matt

VERA, the Virtual Environment for Reactor Applications, is the system of physics capabilities being developed and deployed by the Consortium for Advanced Simulation of Light Water Reactors (CASL). CASL was established for the modeling and simulation of commercial nuclear reactors. VERA consists of integrating and interfacing software together with a suite of physics components adapted and/or refactored to simulate relevant physical phenomena in a coupled manner. VERA also includes the software development environment and computational infrastructure needed for these components to be effectively used. We describe the architecture of VERA from both software and numerical perspectives, along with the goalsmore » and constraints that drove major design decisions, and their implications. We explain why VERA is an environment rather than a framework or toolkit, why these distinctions are relevant (particularly for coupled physics applications), and provide an overview of results that demonstrate the use of VERA tools for a variety of challenging applications within the nuclear industry.« less

ACDOS2: an improved neutron-induced dose rate code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lagache, J.C.

1981-06-01

To calculate the expected dose rate from fusion reactors as a function of geometry, composition, and time after shutdown a computer code, ACDOS2, was written, which utilizes up-to-date libraries of cross-sections and radioisotope decay data. ACDOS2 is in ANSI FORTRAN IV, in order to make it readily adaptable elsewhere.

Simulation of a small cold-leg-break experiment at the PMK-2 test facility using the RELAP5 and ATHLET codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ezsoel, G.; Guba, A.; Perneczky, L.

Results of a small-break loss-of-coolant accident experiment, conducted on the PMK-2 integral-type test facility are presented. The experiment simulated a 1% break in the cold leg of a VVER-440-type reactor. The main phenomena of the experiment are discussed, and in the case of selected events, a more detailed interpretation with the help of measured void fraction, obtained by a special measurement device, is given. Two thermohydraulic computer codes, RELAP5 and ATHLET, are used for posttest calculations. The aim of these calculations is to investigate the code capability for modeling natural circulation phenomena in VVER-440-type reactors. Therefore, the results of themore » experiment and both calculations are compared. Both codes predict most of the transient events well, with the exception that RELAP5 fails to predict the dryout period in the core. In the experiment, the hot- and cold-leg loop-seal clearing is accompanied by natural circulation instabilities, which can be explained by means of the ATHLET calculation.« less

Investigation of the DSMC Approach for Ion/neutral Species in Modeling Low Pressure Plasma Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Deng Hao; Li, Z.; Levin, D.

2011-05-20

Low pressure plasma reactors are important tools for ionized metal physical vapor deposition (IMPVD), a semiconductor plasma processing technology that is increasingly being applied to deposit Cu seed layers on semiconductor surfaces of trenches and vias with the high aspect ratio (e.g., >5:1). A large fraction of ionized atoms produced by the IMPVD process leads to an anisotropic deposition flux towards the substrate, a feature which is critical for attaining a void-free and uniform fill. Modeling such devices is challenging due to their high plasma density, reactive environment, but low gas pressure. A modular code developed by the Computational Opticalmore » and Discharge Physics Group, the Hybrid Plasma Equipment Model (HPEM), has been successfully applied to the numerical investigations of IMPVD by modeling a hollow cathode magnetron (HCM) device. However, as the development of semiconductor devices progresses towards the lower pressure regime (e.g., <5 mTorr), the breakdown of the continuum assumption limits the application of the fluid model in HPEM and suggests the incorporation of the kinetic method, such as the direct simulation Monte Carlo (DSMC), in the plasma simulation.The DSMC method, which solves the Boltzmann equation of transport, has been successfully applied in modeling micro-fluidic flows in MEMS devices with low Reynolds numbers, a feature shared with the HCM. Modeling of the basic physical and chemical processes for ion/neutral species in plasma have been developed and implemented in DSMC, which include ion particle motion due to the Lorentz force, electron impact reactions, charge exchange reactions, and charge recombination at the surface. The heating of neutrals due to collisions with ions and the heating of ions due to the electrostatic field will be shown to be captured by the DSMC simulations. In this work, DSMC calculations were coupled with the modules from HPEM so that the plasma can be self-consistently solved. Differences in the Ar results, the dominant species in the reactor, produced by the DSMC-HPEM coupled simulation will be shown in comparison with the original HPEM results. The effects of the DSMC calculations for ion/neutral species on HPEM plasma simulation will be further analyzed.« less

Comparison of SAND-II and FERRET

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wootan, D.W.; Schmittroth, F.

1981-01-01

A comparison was made of the advantages and disadvantages of two codes, SAND-II and FERRET, for determining the neutron flux spectrum and uncertainty from experimental dosimeter measurements as anticipated in the FFTF Reactor Characterization Program. This comparison involved an examination of the methodology and the operational performance of each code. The merits of each code were identified with respect to theoretical basis, directness of method, solution uniqueness, subjective influences, and sensitivity to various input parameters.

System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

NASA Astrophysics Data System (ADS)

Porter, Ian Edward

A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several additional fuels will also be analyzed, including uranium nitride (UN), uranium carbide (UC) and uranium silicide (U3Si2). Focusing on the system response in an accident scenario, an emphasis is placed on the fracture mechanics of the ceramic cladding by design the fuel rods to eliminate pellet cladding mechanical interaction (PCMI). The time to failure and how much of the fuel in the reactor fails with an advanced fuel design will be analyzed and compared to the current UO2/Zircaloy design using a full scale reactor model.

Multi-Group Formulation of the Temperature-Dependent Resonance Scattering Model and its Impact on Reactor Core Parameters

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ghrayeb, Shadi Z.; Ougouag, Abderrafi M.; Ouisloumen, Mohamed

2014-01-01

A multi-group formulation for the exact neutron elastic scattering kernel is developed. It incorporates the neutron up-scattering effects, stemming from lattice atoms thermal motion and accounts for it within the resulting effective nuclear cross-section data. The effects pertain essentially to resonant scattering off of heavy nuclei. The formulation, implemented into a standalone code, produces effective nuclear scattering data that are then supplied directly into the DRAGON lattice physics code where the effects on Doppler Reactivity and neutron flux are demonstrated. The correct accounting for the crystal lattice effects influences the estimated values for the probability of neutron absorption and scattering,more » which in turn affect the estimation of core reactivity and burnup characteristics. The results show an increase in values of Doppler temperature feedback coefficients up to -10% for UOX and MOX LWR fuels compared to the corresponding values derived using the traditional asymptotic elastic scattering kernel. This paper also summarizes the results done on this topic to date.« less

IMPLEMENTATION AND VALIDATION OF A FULLY IMPLICIT ACCUMULATOR MODEL IN RELAP-7

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhao, Haihua; Zou, Ling; Zhang, Hongbin

2016-01-01

This paper presents the implementation and validation of an accumulator model in RELAP-7 under the framework of preconditioned Jacobian free Newton Krylov (JFNK) method, based on the similar model used in RELAP5. RELAP-7 is a new nuclear reactor system safety analysis code being developed at the Idaho National Laboratory (INL). RELAP-7 is a fully implicit system code. The JFNK and preconditioning methods used in RELAP-7 is briefly discussed. The slightly modified accumulator model is summarized for completeness. The implemented model was validated with LOFT L3-1 test and benchmarked with RELAP5 results. RELAP-7 and RELAP5 had almost identical results for themore » accumulator gas pressure and water level, although there were some minor difference in other parameters such as accumulator gas temperature and tank wall temperature. One advantage of the JFNK method is its easiness to maintain and modify models due to fully separation of numerical methods from physical models. It would be straightforward to extend the current RELAP-7 accumulator model to simulate the advanced accumulator design.« less

Transient analysis of ”2 inch Direct Vessel Injection line break” in SPES-2 facility by using TRACE code

NASA Astrophysics Data System (ADS)

D'Amico, S.; Lombardo, C.; Moscato, I.; Polidori, M.; Vella, G.

2015-11-01

In the past few decades a lot of theoretical and experimental researches have been done to understand the physical phenomena characterizing nuclear accidents. In particular, after the Three Miles Island accident, several reactors have been designed to handle successfully LOCA events. This paper presents a comparison between experimental and numerical results obtained for the “2 inch Direct Vessel Injection line break” in SPES-2. This facility is an integral test facility built in Piacenza at the SIET laboratories and simulating the primary circuit, the relevant parts of the secondary circuits and the passive safety systems typical of the AP600 nuclear power plant. The numerical analysis here presented was performed by using TRACE and CATHARE thermal-hydraulic codes with the purpose of evaluating their prediction capability. The main results show that the TRACE model well predicts the overall behaviour of the plant during the transient, in particular it is able to simulate the principal thermal-hydraulic phenomena related to all passive safety systems. The performance of the presented CATHARE noding has suggested some possible improvements of the model.

Analysis of transient fission gas behaviour in oxide fuel using BISON and TRANSURANUS

NASA Astrophysics Data System (ADS)

Barani, T.; Bruschi, E.; Pizzocri, D.; Pastore, G.; Van Uffelen, P.; Williamson, R. L.; Luzzi, L.

2017-04-01

The modelling of fission gas behaviour is a crucial aspect of nuclear fuel performance analysis in view of the related effects on the thermo-mechanical performance of the fuel rod, which can be particularly significant during transients. In particular, experimental observations indicate that substantial fission gas release (FGR) can occur on a small time scale during transients (burst release). To accurately reproduce the rapid kinetics of the burst release process in fuel performance calculations, a model that accounts for non-diffusional mechanisms such as fuel micro-cracking is needed. In this work, we present and assess a model for transient fission gas behaviour in oxide fuel, which is applied as an extension of conventional diffusion-based models to introduce the burst release effect. The concept and governing equations of the model are presented, and the sensitivity of results to the newly introduced parameters is evaluated through an analytic sensitivity analysis. The model is assessed for application to integral fuel rod analysis by implementation in two structurally different fuel performance codes: BISON (multi-dimensional finite element code) and TRANSURANUS (1.5D code). Model assessment is based on the analysis of 19 light water reactor fuel rod irradiation experiments from the OECD/NEA IFPE (International Fuel Performance Experiments) database, all of which are simulated with both codes. The results point out an improvement in both the quantitative predictions of integral fuel rod FGR and the qualitative representation of the FGR kinetics with the transient model relative to the canonical, purely diffusion-based models of the codes. The overall quantitative improvement of the integral FGR predictions in the two codes is comparable. Moreover, calculated radial profiles of xenon concentration after irradiation are investigated and compared to experimental data, illustrating the underlying representation of the physical mechanisms of burst release.

Experimental investigation of the impulse gas injection into liquid and the use of experimental data for verification of the HYDRA-IBRAE/LM thermohydraulic code

NASA Astrophysics Data System (ADS)

Lobanov, P. D.; Usov, E. V.; Butov, A. A.; Pribaturin, N. A.; Mosunova, N. A.; Strizhov, V. F.; Chukhno, V. I.; Kutlimetov, A. E.

2017-10-01

Experiments with impulse gas injection into model coolants, such as water or the Rose alloy, performed at the Novosibirsk Branch of the Nuclear Safety Institute, Russian Academy of Sciences, are described. The test facility and the experimental conditions are presented in details. The dependence of coolant pressure on the injected gas flow and the time of injection was determined. The purpose of these experiments was to verify the physical models of thermohydraulic codes for calculation of the processes that could occur during the rupture of tubes of a steam generator with heavy liquid metal coolant or during fuel rod failure in water-cooled reactors. The experimental results were used for verification of the HYDRA-IBRAE/LM system thermohydraulic code developed at the Nuclear Safety Institute, Russian Academy of Sciences. The models of gas bubble transportation in a vertical channel that are used in the code are described in detail. A two-phase flow pattern diagram and correlations for prediction of friction of bubbles and slugs as they float up in a vertical channel and of two-phase flow friction factor are presented. Based on the results of simulation of these experiments using the HYDRA-IBRAE/LM code, the arithmetic mean error in predicted pressures was calculated, and the predictions were analyzed considering the uncertainty in the input data, geometry of the test facility, and the error of the empirical correlation. The analysis revealed major factors having a considerable effect on the predictions. The recommendations are given on updating of the experimental results and improvement of the models used in the thermohydraulic code.

I-NERI Quarterly Technical Report (April 1 to June 30, 2005)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chang Oh; Prof. Hee Cheon NO; Prof. John Lee

2005-06-01

The objective of this Korean/United States/laboratory/university collaboration is to develop new advanced computational methods for safety analysis codes for very-high-temperature gas-cooled reactors (VHTGRs) and numerical and experimental validation of these computer codes. This study consists of five tasks for FY-03: (1) development of computational methods for the VHTGR, (2) theoretical modification of aforementioned computer codes for molecular diffusion (RELAP5/ATHENA) and modeling CO and CO2 equilibrium (MELCOR), (3) development of a state-of-the-art methodology for VHTGR neutronic analysis and calculation of accurate power distributions and decay heat deposition rates, (4) reactor cavity cooling system experiment, and (5) graphite oxidation experiment. Second quartermore » of Year 3: (A) Prof. NO and Kim continued Task 1. As a further plant application of GAMMA code, we conducted two analyses: IAEA GT-MHR benchmark calculation for LPCC and air ingress analysis for PMR 600MWt. The GAMMA code shows comparable peak fuel temperature trend to those of other country codes. The analysis results for air ingress show much different trend from that of previous PBR analysis: later onset of natural circulation and less significant rise in graphite temperature. (B) Prof. Park continued Task 2. We have designed new separate effect test device having same heat transfer area and different diameter and total number of U-bands of air cooling pipe. New design has smaller pressure drop in the air cooling pipe than the previous one as designed with larger diameter and less number of U-bands. With the device, additional experiments have been performed to obtain temperature distributions of the water tank, the surface and the center of cooling pipe on axis. The results will be used to optimize the design of SNU-RCCS. (C) Prof. NO continued Task 3. The experimental work of air ingress is going on without any concern: With nuclear graphite IG-110, various kinetic parameters and reaction rates for the C/CO2 reaction were measured. Then, the rates of C/CO2 reaction were compared to the ones of C/O2 reaction. The rate equation for C/CO2 has been developed. (D) INL added models to RELAP5/ATHENA to cacilate the chemical reactions in a VHTR during an air ingress accident. Limited testing of the models indicate that they are calculating a correct special distribution in gas compositions. (E) INL benchmarked NACOK natural circulation data. (F) Professor Lee et al at the University of Michigan (UM) Task 5. The funding was received from the DOE Richland Office at the end of May and the subcontract paperwork was delivered to the UM on the sixth of June. The objective of this task is to develop a state of the art neutronics model for determining power distributions and decay heat deposition rates in a VHTGR core. Our effort during the reporting period covered reactor physics analysis of coated particles and coupled nuclear-thermal-hydraulic (TH) calculations, together with initial calculations for decay heat deposition rates in the core.« less

Creep and Creep-Fatigue Crack Growth at Structural Discontinuities and Welds

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dr. F. W. Brust; Dr. G. M. Wilkowski; Dr. P. Krishnaswamy

2010-01-27

The subsection ASME NH high temperature design procedure does not admit crack-like defects into the structural components. The US NRC identified the lack of treatment of crack growth within NH as a limitation of the code and thus this effort was undertaken. This effort is broken into two parts. Part 1, summarized here, involved examining all high temperature creep-fatigue crack growth codes being used today and from these, the task objective was to choose a methodology that is appropriate for possible implementation within NH. The second part of this task, which has just started, is to develop design rules formore » possible implementation within NH. This second part is a challenge since all codes require step-by-step analysis procedures to be undertaken in order to assess the crack growth and life of the component. Simple rules for design do not exist in any code at present. The codes examined in this effort included R5, RCC-MR (A16), BS 7910, API 579, and ATK (and some lesser known codes). There are several reasons that the capability for assessing cracks in high temperature nuclear components is desirable. These include: (1) Some components that are part of GEN IV reactors may have geometries that have sharp corners - which are essentially cracks. Design of these components within the traditional ASME NH procedure is quite challenging. It is natural to ensure adequate life design by modeling these features as cracks within a creep-fatigue crack growth procedure. (2) Workmanship flaws in welds sometimes occur and are accepted in some ASME code sections. It can be convenient to consider these as flaws when making a design life assessment. (3) Non-destructive Evaluation (NDE) and inspection methods after fabrication are limited in the size of the crack or flaw that can be detected. It is often convenient to perform a life assessment using a flaw of a size that represents the maximum size that can elude detection. (4) Flaws that are observed using in-service detection methods often need to be addressed as plants age. Shutdown inspection intervals can only be designed using creep and creep-fatigue crack growth techniques. (5) The use of crack growth procedures can aid in examining the seriousness of creep damage in structural components. How cracks grow can be used to assess margins on components and lead to further safe operation. After examining the pros and cons of all these methods, the R5 code was chosen as the most up-to-date and validated high temperature creep and creep fatigue code currently used in the world at present. R5 is considered the leader because the code: (1) has well established and validated rules, (2) has a team of experts continually improving and updating it, (3) has software that can be used by designers, (4) extensive validation in many parts with available data from BE resources as well as input from Imperial college's database, and (5) was specifically developed for use in nuclear plants. R5 was specifically developed for use in gas cooled nuclear reactors which operate in the UK and much of the experience is based on materials and temperatures which are experienced in these reactors. If the next generation advanced reactors to be built in the US used these same materials within the same temperature ranges as these reactors, then R5 may be appropriate for consideration of direct implementation within ASME code NH or Section XI. However, until more verification and validation of these creep/fatigue crack growth rules for the specific materials and temperatures to be used in the GEN IV reactors is complete, ASME should consider delaying this implementation. With this in mind, it is this authors opinion that R5 methods are the best available for code use today. The focus of this work was to examine the literature for creep and creep-fatigue crack growth procedures that are well established in codes in other countries and choose a procedure to consider implementation into ASME NH. It is very important to recognize that all creep and creep fatigue crack growth procedures that are part of high temperature design codes are related and very similar. This effort made no attempt to develop a new creep-fatigue crack growth predictive methodology. Rather examination of current procedures was the only goal. The uncertainties in the R5 crack growth methods and recommendations for more work are summarized here also.« less

Reactor Dosimetry Applications Using RAPTOR-M3G:. a New Parallel 3-D Radiation Transport Code

NASA Astrophysics Data System (ADS)

Longoni, Gianluca; Anderson, Stanwood L.

2009-08-01

The numerical solution of the Linearized Boltzmann Equation (LBE) via the Discrete Ordinates method (SN) requires extensive computational resources for large 3-D neutron and gamma transport applications due to the concurrent discretization of the angular, spatial, and energy domains. This paper will discuss the development RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries), a new 3-D parallel radiation transport code, and its application to the calculation of ex-vessel neutron dosimetry responses in the cavity of a commercial 2-loop Pressurized Water Reactor (PWR). RAPTOR-M3G is based domain decomposition algorithms, where the spatial and angular domains are allocated and processed on multi-processor computer architectures. As compared to traditional single-processor applications, this approach reduces the computational load as well as the memory requirement per processor, yielding an efficient solution methodology for large 3-D problems. Measured neutron dosimetry responses in the reactor cavity air gap will be compared to the RAPTOR-M3G predictions. This paper is organized as follows: Section 1 discusses the RAPTOR-M3G methodology; Section 2 describes the 2-loop PWR model and the numerical results obtained. Section 3 addresses the parallel performance of the code, and Section 4 concludes this paper with final remarks and future work.

Calculation and comparison of xenon and samarium reactivities of the HEU, LEU core in the low power research reactor.

PubMed

Dawahra, S; Khattab, K; Saba, G

2015-07-01

Comparative studies for the conversion of the fuel from HEU to LEU in the Miniature Neutron Source Reactor (MNSR) have been performed using the MCNP4C and GETERA codes. The precise calculations of (135)Xe and (149)Sm concentrations and reactivities were carried out and compared during the MNSR operation time and after shutdown for the existing HEU fuel (UAl4-Al, 90% enriched) and the potential LEU fuels (U3Si2-Al, U3Si-Al, U9Mo-Al, 19.75% enriched and UO2, 12.6% enriched) in this paper using the MCNP4C and GETERA codes. It was found that the (135)Xe and (149)Sm reactivities did not reach their equilibrium reactivities during the daily operating time of the reactor. The (149)Sm reactivities could be neglected compared to (135)Xe reactivities during the reactor operating time and after shutdown. The calculations for the UAl4-Al produced the highest (135)Xe reactivity in all the studied fuel group during the reactor operation (0.39 mk) and after the reactor shutdown (0.735 mk), It followed by U3Si-Al (0.34 mk, 0.653 mk), U3Si2-Al (0.33 mk, 0.634 mk), U9Mo-Al (0.3 mk, 0.568 mk) and UO2 (0.24 mk, 0.448 mk) fuels, respectively. Finally, the results showed that the UO2 was the best candidate for fuel conversion to LEU in the MNSR since it gave the lowest (135)Xe reactivity during the reactor operation and after shutdown. Copyright © 2015 Elsevier Ltd. All rights reserved.

Two-phase reduced gravity experiments for a space reactor design

NASA Technical Reports Server (NTRS)

Antoniak, Zenen I.

1987-01-01

Future space missions researchers envision using large nuclear reactors with either a single or a two-phase alkali-metal working fluid. The design and analysis of such reactors require state-of-the-art computer codes that can properly treat alkali-metal flow and heat transfer in a reduced-gravity environment. New flow regime maps, models, and correlations are required if the codes are to be successfully applied to reduced-gravity flow and heat transfer. General plans are put forth for the reduced-gravity experiments which will have to be performed, at NASA facilities, with benign fluids. Data from the reduced-gravity experiments with innocuous fluids are to be combined with normal gravity data from two-phase alkali-metal experiments. Because these reduced-gravity experiments will be very basic, and will employ small test loops of simple geometry, a large measure of commonality exists between them and experiments planned by other organizations. It is recommended that a committee be formed to coordinate all ongoing and planned reduced gravity flow experiments.

CFD Analysis of Coolant Flow in VVER-440 Fuel Assemblies with the Code ANSYS CFX 10.0

DOE Office of Scientific and Technical Information (OSTI.GOV)

Toth, Sandor; Legradi, Gabor; Aszodi, Attila

2006-07-01

From the aspect of planning the power upgrading of nuclear reactors - including the VVER-440 type reactor - it is essential to get to know the flow field in the fuel assembly. For this purpose we have developed models of the fuel assembly of the VVER-440 reactor using the ANSYS CFX 10.0 CFD code. At first a 240 mm long part of a 60 degrees segment of the fuel pin bundle was modelled. Implementing this model a sensitivity study on the appropriate meshing was performed. Based on the development of the above described model, further models were developed: a 960more » mm long part of a 60-degree-segment and a full length part (2420 mm) of the fuel pin bundle segment. The calculations were run using constant coolant properties and several turbulence models. The impacts of choosing different turbulence models were investigated. The results of the above-mentioned investigations are presented in this paper. (authors)« less

Design of the DEMO Fusion Reactor Following ITER.

PubMed

Garabedian, Paul R; McFadden, Geoffrey B

2009-01-01

Runs of the NSTAB nonlinear stability code show there are many three-dimensional (3D) solutions of the advanced tokamak problem subject to axially symmetric boundary conditions. These numerical simulations based on mathematical equations in conservation form predict that the ITER international tokamak project will encounter persistent disruptions and edge localized mode (ELMS) crashes. Test particle runs of the TRAN transport code suggest that for quasineutrality to prevail in tokamaks a certain minimum level of 3D asymmetry of the magnetic spectrum is required which is comparable to that found in quasiaxially symmetric (QAS) stellarators. The computational theory suggests that a QAS stellarator with two field periods and proportions like those of ITER is a good candidate for a fusion reactor. For a demonstration reactor (DEMO) we seek an experiment that combines the best features of ITER, with a system of QAS coils providing external rotational transform, which is a measure of the poloidal field. We have discovered a configuration with unusually good quasisymmetry that is ideal for this task.

Design of the DEMO Fusion Reactor Following ITER

PubMed Central

Garabedian, Paul R.; McFadden, Geoffrey B.

2009-01-01

Runs of the NSTAB nonlinear stability code show there are many three-dimensional (3D) solutions of the advanced tokamak problem subject to axially symmetric boundary conditions. These numerical simulations based on mathematical equations in conservation form predict that the ITER international tokamak project will encounter persistent disruptions and edge localized mode (ELMS) crashes. Test particle runs of the TRAN transport code suggest that for quasineutrality to prevail in tokamaks a certain minimum level of 3D asymmetry of the magnetic spectrum is required which is comparable to that found in quasiaxially symmetric (QAS) stellarators. The computational theory suggests that a QAS stellarator with two field periods and proportions like those of ITER is a good candidate for a fusion reactor. For a demonstration reactor (DEMO) we seek an experiment that combines the best features of ITER, with a system of QAS coils providing external rotational transform, which is a measure of the poloidal field. We have discovered a configuration with unusually good quasisymmetry that is ideal for this task. PMID:27504224

Convection and chemistry effects in CVD: A 3-D analysis for silicon deposition

NASA Technical Reports Server (NTRS)

Gokoglu, S. A.; Kuczmarski, M. A.; Tsui, P.; Chait, A.

1989-01-01

The computational fluid dynamics code FLUENT has been adopted to simulate the entire rectangular-channel-like (3-D) geometry of an experimental CVD reactor designed for Si deposition. The code incorporated the effects of both homogeneous (gas phase) and heterogeneous (surface) chemistry with finite reaction rates of important species existing in silane dissociation. The experiments were designed to elucidate the effects of gravitationally-induced buoyancy-driven convection flows on the quality of the grown Si films. This goal is accomplished by contrasting the results obtained from a carrier gas mixture of H2/Ar with the ones obtained from the same molar mixture ratio of H2/He, without any accompanying change in the chemistry. Computationally, these cases are simulated in the terrestrial gravitational field and in the absence of gravity. The numerical results compare favorably with experiments. Powerful computational tools provide invaluable insights into the complex physicochemical phenomena taking place in CVD reactors. Such information is essential for the improved design and optimization of future CVD reactors.

TRAC-P1: an advanced best estimate computer program for PWR LOCA analysis. I. Methods, models, user information, and programming details

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1978-05-01

The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos Scientific Laboratory (LASL) to provide an advanced ''best estimate'' predictive capability for the analysis of postulated accidents in light water reactors (LWRs). TRAC-Pl provides this analysis capability for pressurized water reactors (PWRs) and for a wide variety of thermal-hydraulic experimental facilities. It features a three-dimensional treatment of the pressure vessel and associated internals; two-phase nonequilibrium hydrodynamics models; flow-regime-dependent constitutive equation treatment; reflood tracking capability for both bottom flood and falling film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions.more » The TRAC-Pl User's Manual is composed of two separate volumes. Volume I gives a description of the thermal-hydraulic models and numerical solution methods used in the code. Detailed programming and user information is also provided. Volume II presents the results of the developmental verification calculations.« less

Comparison of UWCC MOX fuel measurements to MCNP-REN calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Abhold, M.; Baker, M.; Jie, R.

1998-12-31

The development of neutron coincidence counting has greatly improved the accuracy and versatility of neutron-based techniques to assay fissile materials. Today, the shift register analyzer connected to either a passive or active neutron detector is widely used by both domestic and international safeguards organizations. The continued development of these techniques and detectors makes extensive use of the predictions of detector response through the use of Monte Carlo techniques in conjunction with the point reactor model. Unfortunately, the point reactor model, as it is currently used, fails to accurately predict detector response in highly multiplying mediums such as mixed-oxide (MOX) lightmore » water reactor fuel assemblies. For this reason, efforts have been made to modify the currently used Monte Carlo codes and to develop new analytical methods so that this model is not required to predict detector response. The authors describe their efforts to modify a widely used Monte Carlo code for this purpose and also compare calculational results with experimental measurements.« less

Code Development and Assessment for Reactor Outage Thermal-Hydraulic and Safety Analysis - Midloop Operation with Loss of Residual Heat Removal

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liang, Thomas K.S.; Ko, F.-K

Although only a few percent of residual power remains during plant outages, the associated risk of core uncovery and corresponding fuel overheating has been identified to be relatively high, particularly under midloop operation (MLO) in pressurized water reactors. However, to analyze the system behavior during outages, the tools currently available, such as RELAP5, RETRAN, etc., cannot easily perform the task. Therefore, a medium-sized program aiming at reactor outage simulation and evaluation, such as MLO with the loss of residual heat removal (RHR), was developed. All important thermal-hydraulic processes involved during MLO with the loss of RHR will be properly simulatedmore » by the newly developed reactor outage simulation and evaluation (ROSE) code. Important processes during MLO with loss of RHR involve a pressurizer insurge caused by the hot-leg flooding, reflux condensation, liquid holdup inside the steam generator, loop-seal clearance, core-level depression, etc. Since the accuracy of the pressure distribution from the classical nodal momentum approach will be degraded when the system is stratified and under atmospheric pressure, the two-region approach with a modified two-fluid model will be the theoretical basis of the new program to analyze the nuclear steam supply system during plant outages. To verify the analytical model in the first step, posttest calculations against the closed integral midloop experiments with loss of RHR were performed. The excellent simulation capacity of the ROSE code against the Institute of Nuclear Energy Research Integral System Test Facility (IIST) test data is demonstrated.« less

Parallel computation of multigroup reactivity coefficient using iterative method

NASA Astrophysics Data System (ADS)

Susmikanti, Mike; Dewayatna, Winter

2013-09-01

One of the research activities to support the commercial radioisotope production program is a safety research target irradiation FPM (Fission Product Molybdenum). FPM targets form a tube made of stainless steel in which the nuclear degrees of superimposed high-enriched uranium. FPM irradiation tube is intended to obtain fission. The fission material widely used in the form of kits in the world of nuclear medicine. Irradiation FPM tube reactor core would interfere with performance. One of the disorders comes from changes in flux or reactivity. It is necessary to study a method for calculating safety terrace ongoing configuration changes during the life of the reactor, making the code faster became an absolute necessity. Neutron safety margin for the research reactor can be reused without modification to the calculation of the reactivity of the reactor, so that is an advantage of using perturbation method. The criticality and flux in multigroup diffusion model was calculate at various irradiation positions in some uranium content. This model has a complex computation. Several parallel algorithms with iterative method have been developed for the sparse and big matrix solution. The Black-Red Gauss Seidel Iteration and the power iteration parallel method can be used to solve multigroup diffusion equation system and calculated the criticality and reactivity coeficient. This research was developed code for reactivity calculation which used one of safety analysis with parallel processing. It can be done more quickly and efficiently by utilizing the parallel processing in the multicore computer. This code was applied for the safety limits calculation of irradiated targets FPM with increment Uranium.

CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kotas, J.F.; Stroh, K.R.

1983-01-01

The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident thatmore » simulates a control-rod withdrawal at full power.« less

Development of Cross Section Library and Application Programming Interface (API)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, C. H.; Marin-Lafleche, A.; Smith, M. A.

2014-04-09

The goal of NEAMS neutronics is to develop a high-fidelity deterministic neutron transport code termed PROTEUS for use on all reactor types of interest, but focused primarily on sodium-cooled fast reactors. While PROTEUS-SN has demonstrated good accuracy for homogeneous fast reactor problems and partially heterogeneous fast reactor problems, the simulation results were not satisfactory when applied on fully heterogeneous thermal problems like the Advanced Test Reactor (ATR). This is mainly attributed to the quality of cross section data for heterogeneous geometries since the conventional cross section generation approach does not work accurately for such irregular and complex geometries. Therefore, onemore » of the NEAMS neutronics tasks since FY12 has been the development of a procedure to generate appropriate cross sections for a heterogeneous geometry core.« less

Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

2012-06-06

Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spentmore » fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.« less

Volume accumulator design analysis computer codes

NASA Technical Reports Server (NTRS)

Whitaker, W. D.; Shimazaki, T. T.

1973-01-01

The computer codes, VANEP and VANES, were written and used to aid in the design and performance calculation of the volume accumulator units (VAU) for the 5-kwe reactor thermoelectric system. VANEP computes the VAU design which meets the primary coolant loop VAU volume and pressure performance requirements. VANES computes the performance of the VAU design, determined from the VANEP code, at the conditions of the secondary coolant loop. The codes can also compute the performance characteristics of the VAU's under conditions of possible modes of failure which still permit continued system operation.

Physics of reactor safety. Quarterly report, January--March 1977. [LMFBR

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1977-06-01

This report summarizes work done on reactor safety, Monte Carlo analysis of safety-related critical assembly experiments, and planning of DEMI safety-related critical experiments. Work on reactor core thermal-hydraulics is also included.

NGNP Data Management and Analysis System Analysis and Web Delivery Capabilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cynthia D. Gentillon

2010-09-01

Projects for the Very High Temperature Reactor Technology Development Office provide data in support of Nuclear Regulatory Commission licensing of the very high temperature reactor. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high-temperature and high-fluence environments. In addition, thermal-hydraulic experiments are conducted to validate codes used to assess reactor safety. The Very High Temperature Reactor Technology Development Office has established the NGNP Data Management and Analysis System (NDMAS) at the Idaho National Laboratory to ensure that very high temperature reactor data are (1) qualified for use, (2) stored in amore » readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the third NDMAS objective. It describes capabilities for displaying the data in meaningful ways and for data analysis to identify useful relationships among the measured quantities.« less

Conceptual design of a fast-ignition laser fusion reactor based on a dry wall chamber

NASA Astrophysics Data System (ADS)

Ogawa, Y.; Goto, T.; Okano, K.; Asaoka, Y.; Hiwatari, R.; Someya, Y.

2008-05-01

The fast ignition is quite attractive for a compact laser fusion reactor, because a sufficiently high pellet gain is available with a small input energy. We designed an inertial fusion reactor based on Fast-ignition Advanced Laser fusion reactor CONcept, called FALCON-D, where a dry wall is employed for a chamber wall. A simple point model shows that the pellet gain G~100 is available with laser energies of 350kJ for implosion, 50kJ for heating. This results in the fusion yield of 40 MJ in one shot. By increasing the repetition rate up to 30 Hz, the fusion power of 1.2 GWth becomes available. Plant system analysis shows the net electric power to be about 0.4 GWe In the fast ignition it is available to employ a low aspect ratio pellet, which is favorable for the stability during the implosion phase. Here the pellet aspect ratio is reduced to be 2 ~ 4, and the optimization of the pulse shape for the implosion laser are carried out by using the 1-D hydrodynamic simulation code ILESTA-1D. A ferritic steel with a tungsten armour is employed for the chamber wall. The feasibility of this dry wall concept is studied from various engineering aspects such as surface melting, physical and chemical sputtering, blistering and exfoliation by helium retention, and thermo-mechanical fatigue, and it is found that blistering and exfoliation due to the helium retention and fatigue failure due to cyclic thermal load are major concerns. The cost analysis shows that the construction cost is moderate but the cost of electricity is slightly expensive.

SIGACE Code for Generating High-Temperature ACE Files; Validation and Benchmarking

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sharma, Amit R.; Ganesan, S.; Trkov, A.

2005-05-24

A code named SIGACE has been developed as a tool for MCNP users within the scope of a research contract awarded by the Nuclear Data Section of the International Atomic Energy Agency (IAEA) (Ref: 302-F4-IND-11566 B5-IND-29641). A new recipe has been evolved for generating high-temperature ACE files for use with the MCNP code. Under this scheme the low-temperature ACE file is first converted to an ENDF formatted file using the ACELST code and then Doppler broadened, essentially limited to the data in the resolved resonance region, to any desired higher temperature using SIGMA1. The SIGACE code then generates a high-temperaturemore » ACE file for use with the MCNP code. A thinning routine has also been introduced in the SIGACE code for reducing the size of the ACE files. The SIGACE code and the recipe for generating ACE files at higher temperatures has been applied to the SEFOR fast reactor benchmark problem (sodium-cooled fast reactor benchmark described in ENDF-202/BNL-19302, 1974 document). The calculated Doppler coefficient is in good agreement with the experimental value. A similar calculation using ACE files generated directly with the NJOY system also agrees with our SIGACE computed results. The SIGACE code and the recipe is further applied to study the numerical benchmark configuration of selected idealized PWR pin cell configurations with five different fuel enrichments as reported by Mosteller and Eisenhart. The SIGACE code that has been tested with several FENDL/MC files will be available, free of cost, upon request, from the Nuclear Data Section of the IAEA.« less

Low Energy Neutrino Physics at the Kuo-Sheng Reactor Laboratory in Taiwan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lin, S.-T.

2006-11-17

A laboratory has been constructed by the TEXONO Collaboration at the Kuo-Sheng Reactor Power Plant in Taiwan to study low energy neutrino physics. A limit on the neutrino magnetic moment of {mu}{nu}({nu}-bare) < 7.2 x 10-11 {mu}B at 90% confidence level has been achieved from measurements with a high-purity germanium detector, as well as the electron neutrinos ({nu}{sub e}) produced from nuclear power reactors has been studied. Other research program at Kuo-Sheng are surveyed.

The Paucity Problem: Where Have All the Space Reactor Experiments Gone?

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, John D.; Marshall, Margaret A.

2016-10-01

The Handbooks of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) together contain a plethora of documented and evaluated experiments essential in the validation of nuclear data, neutronics codes, and modeling of various nuclear systems. Unfortunately, only a minute selection of handbook data (twelve evaluations) are of actual experimental facilities and mockups designed specifically for space nuclear research. There is a paucity problem, such that the multitude of space nuclear experimental activities performed in the past several decades have yet to be recovered and made available in such detail that themore » international community could benefit from these valuable historical research efforts. Those experiments represent extensive investments in infrastructure, expertise, and cost, as well as constitute significantly valuable resources of data supporting past, present, and future research activities. The ICSBEP and IRPhEP were established to identify and verify comprehensive sets of benchmark data; evaluate the data, including quantification of biases and uncertainties; compile the data and calculations in a standardized format; and formally document the effort into a single source of verified benchmark data. See full abstract in attached document.« less

Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

NASA Astrophysics Data System (ADS)

Cisneros, Anselmo Tomas, Jr.

The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP and PEBBED for a high temperature gas cooled pebble bed reactor. Three parametric studies were performed for exploring the design space of the PB-FHR---to select a fuel design for the PB-FHR] to select a core configuration; and to optimize the PB-FHR design. These parametric studies investigated trends in the dependence of important reactor performance parameters such as burnup, temperature reactivity feedback, radiation damage, etc on the reactor design variables and attempted to understand the underlying reactor physics responsible for these trends. A pebble fuel parametric study determined that pebble fuel should be designed with a carbon to heavy metal ratio (C/HM) less than 400 to maintain negative coolant temperature reactivity coefficients. Seed and thorium blanket-, seed and inert pebble reflector- and seed only core configurations were investigated for annular FHR PBRs---the C/HM of the blanket pebbles and discharge burnup of the thorium blanket pebbles were additional design variable for core configurations with thorium blankets. Either a thorium blanket or graphite pebble reflector is required to shield the outer graphite reflector enough to extend its service lifetime to 60 EFPY. The fuel fabrication costs and long cycle lengths of the thorium blanket fuel limit the potential economic advantages of using a thorium blanket. Therefore, the seed and pebble reflector core configuration was adopted as the baseline core configuration. Multi-objective optimization with respect to economics was performed for the PB-FHR accounting for safety and other physical design constraints derived from the high-level safety regulatory criteria. These physical constraints were applied along in a design tool, Nuclear Application Value Estimator, that evaluated a simplified cash flow economics model based on estimates of reactor performance parameters calculated using correlations based on the results of parametric design studies for a specific PB-FHR design and a set of economic assumptions about the electricity market to evaluate the economic implications of design decisions. The optimal PB-FHR design---Mark 1 PB-FHR---is described along with a detailed summary of its performance characteristics including: the burnup, the burnup evolution, temperature reactivity coefficients, the power distribution, radiation damage distributions, control element worths, decay heat curves and tritium production rates. The Mk1 PB-FHR satisfies the PB-FHR safety criteria. The fuel, moderator (pebble core, pebble shell, graphite matrix, TRISO layers) and coolant have global negative temperature reactivity coefficients and the fuel temperatures are well within their limits.

The Effect of Cold Work on Properties of Alloy 617

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wright, Richard

2014-08-01

Alloy 617 is approved for non-nuclear construction in the ASME Boiler and Pressure Vessel Code Section I and Section VIII, but is not currently qualified for nuclear use in ASME Code Section III. A draft Code Case was submitted in 1992 to qualify the alloy for nuclear service but efforts were stopped before the approval process was completed.1 Renewed interest in high temperature nuclear reactors has resulted in a new effort to qualify Alloy 617 for use in nuclear pressure vessels. The mechanical and physical properties of Alloy 617 were extensively characterized for the VHTR programs in the 1980’s andmore » incorporated into the 1992 draft Code Case. Recently, the properties of modern heats of the alloy that incorporate an additional processing step, electro-slag re-melting, have been characterized both to confirm that the properties of contemporary material are consistent with those in the historical record and to increase the available database. A number of potential issues that were identified as requiring further consideration prior to the withdrawal of the 1992 Code Case are also being re-examined in the current R&D program. Code Cases are again being developed to allow use of Alloy 617 for nuclear design within the rules of the ASME Boiler and Pressure Vessel Code. In general the Code defines two temperature ranges for nuclear design with austenitic and nickel based alloys. Below 427°C (800°F) time dependent behavior is not considered, while above this temperature creep and creep-fatigue are considered to be the dominant life-limiting deformation modes. There is a corresponding differentiation in the treatment of the potential for effects associated with cold work. Below 427°C the principal issue is the relationship between the level of cold work and the propensity for stress corrosion cracking and above that temperature the primary concern is the impact of cold work on creep-rupture behavior.« less